阅读说明：本技术 用于基于中子活化进行多元素分析的方法和装置、以及用途 (Method and device for neutron activation based multi-element analysis, and use ) 是由 卡伊·卡基 约翰·克特勒 安德里亚斯·哈维尼斯 于 2018-05-28 设计创作，主要内容包括：本发明涉及一种用于基于中子活化进行多元素分析的方法,该方法包括以下步骤：产生能量在10keV至20MeV范围内的快中子；用这些中子辐照样品(1)；测量由经辐照的样品发射的伽马辐射,以便确定该样品中的至少一种元素；其中,根据本发明,该样品是以非脉冲且连续的方式辐照的,在辐照期间进行测量,至少瞬发伽马辐射或瞬发伽马辐射和延迟伽马辐射两者被测量和评估,以便确定该至少一种元素,该样品被划分成各个分区并且使用准直器进行测量,并且在该样品(1)的相关分区(P1,P2,Pn)内以空间分辨和能量分辨的方式确定中子流。因此扩展了分析,提供了一种灵活的方法。本发明还涉及相应的装置以及探测器单元在多元素分析中的用途。(The invention relates to a method for neutron activation based multi-element analysis, comprising the following steps: generating fast neutrons with energy in the range of 10keV to 20 MeV; irradiating the sample (1) with these neutrons; measuring gamma radiation emitted by the irradiated sample to determine at least one element in the sample; wherein, according to the invention, the sample is irradiated in a non-pulsed and continuous manner, measurements are taken during the irradiation, at least prompt gamma radiation or both prompt and delayed gamma radiation are measured and evaluated in order to determine the at least one element, the sample is divided into sections and the measurements are taken using collimators, and the neutron flow is determined in a spatially and energy-resolved manner within the relevant sections (P1, P2, Pn) of the sample (1). Thus extending the analysis and providing a flexible approach. The invention also relates to a corresponding device and to the use of a detector unit in a multi-element analysis.)

1. A method for neutron activation based multi-element analysis, the method comprising the steps of:

generating fast neutrons with energy in the range of 10keV to 20 MeV;

irradiating the sample (1) with these neutrons;

measuring gamma radiation emitted by the irradiated sample to determine at least one element in the sample;

characterized in that the sample is continuously irradiated in a non-pulsed manner, wherein the measurement is carried out during such irradiation, wherein, for determining the at least one element, at least prompt gamma radiation or both prompt gamma radiation and delayed gamma radiation from such continuous neutron irradiation is measured and evaluated, wherein the sample (1) is subdivided into individual sections (P1, P2, Pn), and the measurement is carried out using a collimator (17) in relation to the respective sections (P1, P2, Pn), and wherein the determination comprises an evaluation of the measured gamma radiation, wherein the evaluation comprises: spatially and energy-resolved determination of the neutron flux in the respective sub-region (P1, P2, Pn) of the sample (1) is carried out.

2. The method of claim 1, wherein the irradiating and measuring are performed over a time period of at least one millisecond or at least one second; and/or wherein the neutrons produced have a neutron energy value of 2.45MeV or at least one neutron energy value selected from the group consisting of: 2.45MeV, 14.1 MeV; and/or wherein the neutrons produced have neutron energies of at least one value in an energy range of 10keV to 20MeV or 10keV to 10 MeV.

3. The method of any one of the preceding claims, wherein only delayed gamma radiation from continuous neutron irradiation is measured and evaluated at least intermittently in order to determine the at least one element; or wherein such measurement or determination is carried out individually with respect to individual partitions (P1, P2, Pn) of the sample, which partitions are predefined manually or automatically by collimation or which can be predefined manually or automatically by collimation.

4. The method of any one of the preceding claims, wherein, for determining the at least one element, the gamma radiation emitted by the sample (1) is measured in an energy-resolved manner by determining a photopeak count rate, wherein such determination comprises an energy-resolved evaluation of the measured gamma radiation from the gamma energy spectra of the respective partitions; and/or wherein such measurement/evaluation comprises an energy-resolved measurement/evaluation of the intensity of gamma radiation emitted by the sample (1).

5. The method of any one of the preceding claims, wherein such assessment comprises: correlating at least one photopeak in the count rate-energy map with an element of the sample (1) based on its energy; or wherein such evaluating further comprises: after subtracting the background signal from the net area of a/the photopeak in the count rate-energy plot caused by the element, the mass fraction of the at least one element in the sample is quantified by means of the component of the at least one element contained in the sample being evaluated.

6. The method of any one of the preceding claims, wherein the evaluation is carried out on the assumption of a homogeneous distribution of masses and/or elements in the respective partition of the sample (1); and/or wherein the neutron flux within the respective section of the sample (1) is calculated based on a diffusion approximation of the linear boltzmann equation, in particular based on the following relation:

and/or wherein the neutron spectrum is calculated within the sample (1), within the respective section (P1, Pn) of the sample, in particular in a spatially and/or energy-resolved manner, in particular on the basis of the following relation:

7. the method of any one of the preceding claims, wherein such assessment comprises: by calculating the neutron flux and the neutron energy spectrum in an approximation method, the energy-dependent photopeak efficiency and the neutron flux and the neutron energy spectrum within the individual sections (P1, P2, Pn) of the sample are calculated in each case on the basis of the following relationships:

and/or wherein, during such evaluation, a plurality of gamma energies of at least one element in a respective partition (P1, P2, Pn) of the sample (1) is analyzed accordingly when quantifying the mass fraction of the respective element in the respective partition, said analysis being based on the following relation:

8. the method as claimed in any of the preceding claims, wherein the method is performed on the basis of input variables of neutron source intensity, sample geometry and sample mass, in particular exclusively on the basis of the three input variables, wherein the method is iteratively performed in each case with respect to individual elements and/or with respect to the respective section (P1, P2, Pn) of the sample (1) and/or with respect to the complete composition of the sample (1); and/or wherein the method is performed in an automated manner by evaluating the measured gamma radiation based on parameters determined purely numerically in addition to the three parameters of neutron source intensity, sample geometry and sample quality during such irradiation; and/or wherein at least one measurement from the following set of measurements is performed to characterize the sample: transmission measurement, sample weighing, optical detection of the sample geometry.

9. The method of one of the preceding claims, wherein the spatially and energy-resolved determination of the neutron flux, in particular of the total neutron flux, of the respective sub-region is carried out within the sample chamber and outside the sample, in particular by means of at least one neutron detector (28; 28A, 28B, 28C, 28D) arranged within the sample chamber.

10. Use of a detector unit (16) comprising at least one detector (16A, 16B) according to the method of any of the preceding method claims in a neutron activation based multi-element analysis of a sample (1), configured to continuously measure both prompt and delayed gamma radiation emitted due to continuous irradiation of the sample by neutrons, wherein also such gamma radiation is measured at least partly continuously and simultaneously with such continuous irradiation, wherein the field of view of the detector unit (16) is limited to a respective partition of the sample (1) by at least one collimator (17).

11. A control device (20) configured to activate at least one neutron generator (11) of an apparatus for multi-element analysis based on neutron activation according to the method of any of the preceding method claims, wherein the neutron generator (11) is configured to generate fast neutrons with an energy in the range of 10keV to 20MeV, wherein the control device (20) is configured to activate the neutron generator to generate neutrons and irradiate a sample (1) in a non-pulsed, continuous manner, and wherein the control device is further configured to activate at least one detector (16, 16A, 16B) to continuously and/or intermittently measure gamma radiation emitted by a respective partition (P1, P2, Pn) of the sample simultaneously with such irradiation, and to activate the detector (16A) by means of at least one collimator (17), 16B) is limited to the respective partition (P1, P2, Pn) of the sample (1).

12. A computer program product (30) configured to perform a multi-element analysis based on neutron activation according to the method of any of the preceding method claims and to determine at least one element in a sample (1) which is irradiated with neutrons in a non-pulse continuous manner by evaluating at least prompt gamma radiation or prompt and delayed gamma radiation emitted by the sample based on an energy-dependent photopeak efficiency as a function of the composition of the sample (1) and neutron flux and neutron energy spectra within the respective partition (P1, P2, Pn) of the sample, and further configured to evaluate the measured gamma radiation in a partition-collimated manner by a plurality of gamma energies of the at least one element, the plurality of gamma energies being respectively within the respective partition (P1, p2, Pn) when quantifying the mass fraction of the respective elements of the respective partition (P1, P2, Pn) based on the net photopeak count rate recorded during such multi-element analysis, in particular based on the following relation:

13. an apparatus (10) configured to perform neutron activation based multi-element analysis according to the method of any of the preceding method claims, the apparatus comprising:

a neutron generator (11) configured to generate fast neutrons;

a sample chamber (15) and a sample holder (14) arranged therein;

a detector unit (16) comprising at least one detector (16A, 16B) configured to measure gamma radiation emitted by the irradiated sample in order to determine at least one element in the sample;

characterized in that the apparatus (10) is configured to irradiate a/the sample (1) in a non-pulsed continuous manner, wherein the apparatus comprises at least one collimator (17) limiting the field of view of the detector (16A, 16B) to respective sections (P1, P2, Pn) of the sample (1), and is configured to subdivide the sample (1) into individual sections (P1, P2, Pn), and is further configured to measure at least prompt gamma radiation or both prompt and delayed gamma radiation emitted by the continuously irradiated sample during such irradiation with respect to the respective sections (P1, P2, Pn) of the sample (1), the apparatus further comprising control means (20) configured to perform automatic continuous irradiation and configured to control/adjust automatic measurement of the neutron irradiation continuously applied during such irradiation, wherein the apparatus (10) is further configured to determine neutron flux within respective partitions (P1, P2, Pn) of the sample (1) in a spatially and energy-resolved manner and to evaluate measurements of the partitions (P1, P2, Pn) by quantifying a mass fraction of the at least one element in the sample (1).

14. The apparatus of the preceding apparatus claim, wherein the neutron generator (11) comprises a neutron source (11.1) configured to fuse deuterons, in particular using deuterium gas as fuel; and/or wherein the device comprises at least one component attenuating a background signal of the device, said component being selected from the group of: at least one collimator (17) made of lead or bismuth, said collimator limiting the field of view of the detector to a respective sector of the sample; and/or a moderating chamber (12) made of graphite; and/or a shield (19) made of boron-containing polyethylene; and/or the sample chamber (15) and/or the sample carrier (14), each being at least partially made of graphite or perfluorinated plastic or beryllium.

15. The apparatus of any one of the preceding apparatus claims, further comprising a computer program product (30) or a data storage (21) with the computer program product, wherein the computer program product is configured to determine the at least one element in the sample (1) by evaluating the measured gamma radiation based on an energy-dependent photopeak efficiency and neutron flux and neutron spectrum within a respective partition (P1, P2, Pn) of the sample; and/or further comprising a rotation and/or lifting means (18) configured to displace a certain/the sample carrier (14) or the sample in a translational and/or rotational manner, in particular a rotation and/or lifting means decoupled from a certain/the sample chamber (15) of the apparatus; and/or further comprising at least two detectors (16A, 16B), in particular symmetrically arranged with respect to the neutron generator (11) or with respect to at least one neutron source (11.1) of the apparatus.

16. An apparatus (10) configured to perform a neutron activation based multi-element analysis according to the method of any of the preceding method claims.

17. Use of a neutron generator (11) configured to fuse deuterons for the purpose of generating fast neutrons for the continuous non-pulsed irradiation of a sample according to the method of any of the preceding method claims, in the multi-element analysis of the sample based on neutron activation.

Technical Field

The present invention relates to a method for neutron activation based multi-element analysis by irradiating a sample with neutrons. Further, the invention relates to a corresponding apparatus comprising at least one collimatable detector. Of particular importance, the invention also relates to the use of the control device or the computer program product thereof. In particular, the invention relates to a method and an apparatus according to the preambles of the respective or alternative independent claims.

Background

In many industrial fields, in particular in the field of hazardous goods or waste or of recycled materials or raw materials, or in the field of quality control of semi-finished or industrial products, the analysis of substances or materials, in particular with regard to their elemental composition, is of great importance. One analysis method carried out previously is the so-called multi-element analysis, by means of which individual elements of a sample are determined without the exact composition of the sample being known beforehand.

The multielement analysis may be performed by neutron activation or, for example, by X-ray fluorescence analysis or mass spectrometry. Previously, multielement analysis by neutron activation was carried out by means of irradiation, according to a certain time specification. In the case of pulsed neutron irradiation, it can be ensured that meaningful measurement results are obtained if, after a certain time window, the prompt gamma radiation is evaluated in the manner of pulsed irradiation. After a waiting time after the end of each neutron pulse, a time window for detecting gamma radiation is started and the time window ends before the next neutron pulse is emitted.

WO 2012/010162 a1 and DE 102010031844 a1 describe a method for the non-destructive elemental analysis of large volumes of samples using neutron radiation and an apparatus for carrying out the method. In the method, the sample is irradiated in pulses by fast neutrons, wherein after a certain time window after the neutron pulse, the gamma radiation emitted by the sample is measured before a new neutron pulse is emitted towards the sample. Here, the measurement method is also based on the following findings: the measurement can be facilitated by a slowing down process and by observing a time window after the corresponding neutron pulse. Due to the time window after the respective neutron pulse, the detection of induced gamma radiation in the detector caused by inelastic interactions may be filtered out and thus may be shielded during the measurement. Prompt gamma radiation is evaluated as gamma radiation.

EP 1882929B 1 and WO 01/07888 a2 also describe methods in which neutrons are radiated onto the sample in pulses and a certain time window is observed after each pulse until prompt gamma radiation emitted by the sample is measured. Similar processes are also described, for example, in EP 0493545B 1 and DE 102007029778B 4.

Neutron activation analysis is also described in the following additional publications: US 2015/0338356 a1, DE 60310118T 2, US 2005/0004763 a1, US 2012/046867 a1, DE 10215070 a1, DE 1236831B.

Disclosure of Invention

It is an object to provide a method and apparatus that can simplify multi-element analysis of a sample by neutron activation. Another object is to construct a method and apparatus for multi-element analysis by neutron activation, so that a wide range of application fields results. The object may also be regarded as providing the user with a simple, fast measuring method which can be applied as independently as possible of the type or size of the sample to be analyzed or the material composition. It is particularly important to provide a non-destructive method which is as flexible as possible and an apparatus for multi-element analysis based on neutron activation which has a very high quality or a very reliable, safe way of measuring and evaluating the emitted gamma radiation even if the sample to be examined is difficult to analyze in terms of elemental composition and sample geometry and/or if it is not desirable to collect parts of the sample to be examined which are destructive or which affect the result.

At least one of these objects is achieved by a method according to claim 1 and by a device according to the alternative independent device claim. Advantageous developments of the invention are specified in the respective dependent claims. Features of the exemplary embodiments described below may be combined with each other as long as this is not explicitly prohibited.

A method for neutron activation based multi-element analysis is performed by the steps of: generating fast neutrons with energy in the range of 10keV to 20 MeV; irradiating the sample with these neutrons; gamma radiation emitted by the irradiated sample is measured to determine at least one element in the sample. According to the invention, it is proposed that the sample is irradiated continuously in a non-pulsed manner, wherein the measurement is carried out during the irradiation independently of the irradiation time (independently of the irradiation time profile), in particular without a time window predetermined by the neutron pulse, in particular simultaneously with the irradiation, in particular continuously during the irradiation, within the same time period as the irradiation.

This may provide a non-destructive method for multi-element analysis based on neutron activation for various types of samples, with high measurement flexibility and robust, reproducible and reliable results. The sample is irradiated continuously with neutrons without separate pulses, for example over a period of several seconds or minutes or hours, wherein the gamma radiation emitted by the sample can be measured simultaneously with the irradiation. It is found here that neutrons can be generated in particular by a generator configured to fuse deuterons (deuterium nuclei), in particular with deuterium gas as gas target or fuel. The invention facilitates measurements and evaluations over a long period of time based on irradiation with comparatively low energy, as a result of which analysis can be carried out very accurately and reproducibly.

In the prior art, pulse irradiation has previously been used in many measurement problems, with a waiting time being required in advance after the corresponding neutron pulse. Heretofore, pulse lengths have typically ranged from ten microseconds (μ s) to several hundred microseconds (ps). In contrast to pulsed irradiation, both prompt and delayed gamma radiation emitted by the sample are measured simultaneously with the irradiation, with the energy resolution of the detector contributing to the subdivision of the prompt and delayed gamma radiation. In contrast to pulsed irradiation, no time window needs to be observed (to date, this is typically at least 5 μ s) before the emitted radiation can be measured/evaluated. There is no need to observe the waiting time before starting to detect gamma radiation. No longer need to coordinate the time with the end of the corresponding neutron pulse; alternatively, irradiation and measurement may be performed continuously. This also allows to reduce the measurement time for analyzing the sample.

The measurement may be performed entirely without using a time window, or alternatively, may be performed partially using a time window. In any case, at least some of the emitted gamma radiation is measured in a manner that has no time window independent of time. The method may include at least intermittently moderating fast neutrons.

The simultaneous measurement of gamma radiation enables a high efficiency of the measuring facility. Here, both prompt and delayed gamma radiation can be measured in the standard mode of operation, with the emphasis on prompt gamma radiation. In particular, the measurement can be carried out simultaneously with the irradiation in a continuous manner, either at the same time specification as the irradiation or in individual time windows independent of the time specification of the irradiation. For example, the irradiation is continuous, but optionally only measured during short time intervals. As a result of the simultaneous measurement and continuous irradiation, no longer any time window has to be taken into account; instead, the measurement and evaluation can be made very flexible and both types of radiation, i.e. prompt gamma radiation and delayed gamma radiation, can be evaluated. The measurement/detection of gamma radiation independent of time during neutron irradiation can be highlighted as a feature of the invention.

According to the prior art, the irradiation is pulsed, requiring a waiting time after the corresponding neutron pulse. Heretofore, pulse lengths have typically ranged from ten microseconds (μ s) to several hundred microseconds (ps). In the measurement arrangements employed heretofore, the background signal immediately after the respective neutron pulse is too high, and therefore the signal-to-noise ratio (SNR) immediately after the neutron pulse is too poor to evaluate the measurement results. Therefore, it is impossible to detect a meaningful gamma radiation energy spectrum. In previous methods, neutrons are provided with a certain time window in many measurement problems in order to be able to measure, in particular, after neutron emission. Typically, this time window is at least 5 μ s. The probability of interaction in the sample increases during this time window, so that measurements can be performed with a sufficiently good signal-to-noise ratio SNR after a certain waiting time (or moderation time) has elapsed after the corresponding neutron pulse. The data acquisition is performed in a time-shifted manner from the neutron pulse.

In contrast, first, the gamma radiation measured and evaluated according to the present method is prompt gamma radiation emitted immediately after a neutron interacts with a sample nucleus. In the case of prompt gamma radiation, the time period until emission is about 10exp-16 to 10exp-12 seconds; this period of time is too short and may be referred to as instantaneous/immediate transmission. In the case of prompt gamma radiation, there is no time offset between the neutron capture and the emission of gamma radiation that is relevant from a measurement perspective. Secondly, delayed gamma radiation is also affected; this delayed gamma radiation is emitted as the activated nuclei decay in a time-shifted manner, according to a characteristic half-life. Depending on the characteristic half-life of the radionuclide formed, after neutron capture, the nuclei will emit delayed gamma radiation with a time shift. In conventional Neutron Activation Analysis (NAA), the cross section of the neutron capture and the half-life of the activated radionuclide have an effect on the emitted radiation. According to the invention, the following two measurement concepts can be associated with each other: on the one hand conventional Neutron Activation Analysis (NAA) and on the other hand prompt gamma NAA (pgnaa). Here, prompt gamma radiation and delayed gamma radiation can be distinguished on the basis of the energy of the gamma radiation (in particular the position of the maximum peak) and the energy resolution of the detector.

Most knowledge can be gained by evaluating prompt gamma radiation, as expected. However, the presence of a large number of elements (e.g. lead) does not provide a good transient signal. Therefore, it is convenient to evaluate both prompt and delayed gamma radiation in many applications or many types of material samples. Optionally, the irradiation can also optionally be carried out at least intermittently in a pulsed manner, so that only delayed gamma radiation is measured. Alternatively, in particular in the case of an analysis with regard to lead, only delayed gamma radiation can be measured independently of the type of irradiation.

In one or more detectors, gamma radiation emitted by the sample is measured in an energy-resolved manner. This produces a measured gamma energy spectrum as a function of energy, which in particular corresponds to a record of the number of events detected in the gamma ray detector. Elements in the sample are identified based on the energy of the gamma radiation. The elemental mass is quantified by means of the measured energy-dependent radiation intensity. After subtracting the background signal, the mass fraction of the element contained in the sample is estimated from the area of the light peak in the gamma spectrum caused by the element. Since, as a rule, the irradiated elements in the sample emit gamma radiation of different energies, all appreciable gamma energies of the elements are taken into account in the analytical evaluation both for the determination of the mass and for the uncertainty analysis with respect to the mass determination. It has been found that an advantage of analyzing based on all evaluable gamma energies of an element is that a broad data base can be used and a plausibility check can be performed. Further, the measurement uncertainty still present can be reduced and the accuracy of the measurement method can be improved.

In the case of multi-element analysis, the analytical evaluation for determining the mass is based in particular on the calculation of the optical peak efficiency from the sample and the energy-dependent gamma emission from the individual sections of the sample, and on the calculation of the neutron energy spectrum and the neutron flux within the sample and within the sections of the sample. Initially, assumptions about the basis composition can be made for these calculations, which are derived from the evaluation of the gamma spectra. It has been found that the results of the multi-element analysis define initial assumptions made a priori for more accurate calculation of the photopeak efficiency, neutron energy spectrum and neutron flux, and can also improve the accuracy of the measurement method, so that the method is preferably performed iteratively with respect to the composition of the sample until the calculated sample composition is stable. Due to this type of analysis procedure, a non-destructive method for multi-element analysis based on neutron activation can be performed automatically, in particular iteratively, wherein only the form and mass of the sample body and the neutron source intensity are required as input parameters. Here, the neutron source intensity may be obtained from a measurement facility or apparatus as a control variable.

The method for neutron activation based multi-element analysis and the apparatus for performing the method allow non-destructive examination of various samples with respect to the composition of matter in a simple manner. Examples of analyzable samples may include: soil samples, ash, water samples, sludge, electronic waste, chemical toxic or radioactive waste. Here, the sample can be analyzed with respect to mass flow in a batch operation or on-line. The samples may be analyzed for quality assurance, target classification, process control and/or validation management, among other purposes.

In contrast to previous methods, the method according to the invention is characterized in particular by the following characteristics: continuous emission of fast neutrons; continuous measurement of gamma energy spectra; collimated measurement of the entire sample or of individual partial volumes (partitions) of the sample; in particular irradiating the sample with 2.45MeV neutrons (<10MeV, relatively low initial energy); evaluating gamma radiation by evaluating the signal of each partition; the evaluation of the analysis for determining the element quality is carried out in particular on the assumption of a simplification that the element quality is distributed uniformly within the subareas; and/or there is rotational and axial displacement of the sample body relative to the probe. After and during fast neutron emission, fast neutrons may be moderated in a moderating chamber, sample chamber, and/or the sample itself until the neutrons are sufficiently thermalized.

In contrast, analysis has heretofore generally been conducted by a method having the following characteristics: performing pulse irradiation by using fast neutrons; measuring the gamma energy spectrum after the respective neutron pulse or within a predefined time interval or time window between individual neutron pulses; the sample is measured as a whole without defining a collimator; in particular, the sample was irradiated with 14.1MeV neutrons (>10 MeV); the sample body is rotated in front of the detector and gamma radiation is measured from the angle of rotation of the irradiated sample; the analytical evaluation for determining the element quality of the inhomogeneous distribution is carried out on the basis of the simplified assumption that the element quality is punctiform. The integrated neutron flux in the sample can be determined by the metal shield of the sample.

The following briefly discusses the terminology used in connection with the present invention.

A shield should preferably be understood to mean a material or unit that surrounds the device or the measurement facility and reduces the ambient gamma and neutron dose rates outside the measurement facility.

"irradiation" should preferably be understood to mean the operation of a neutron generator and the generation and emission of neutrons towards at least one sample, so as to cause the emission of gamma radiation that is characteristic of the elemental composition from the sample.

A detector unit is preferably understood to mean a unit or a component of a measuring facility comprising one or more detectors which carry out high-resolution measurements of gamma radiation emitted by the sample or individual sections of the sample. The corresponding detector may have a range of, for example, 5cm to 10cm in one spatial direction.

A collimator is preferably understood to mean a unit or component of the measuring facility which limits the field of view of the detector to a spatial region with a high probability of detecting gamma radiation. Collimation may also be performed specifically with respect to various sections/partitions of the sample.

A measuring facility is preferably understood to mean a metering facility which generates ionizing radiation for the purpose of multi-element analysis of a sample. In one embodiment, the apparatus described herein may be referred to as a measurement facility.

The moderating chamber is preferably understood to mean a component of the measuring device for moderating neutrons, in particular by graphite, or a component which is at least partially composed of graphite. Depending on the desired type of measurement/evaluation, the moderation may optionally be provided in the sample chamber, and/or may be implemented in a separate moderating chamber.

A neutron generator, which emits fast neutrons (in particular 2.45MeV neutrons, or further neutrons of generally <10 MeV) and which is arranged within the shield, is preferably understood to refer to a component of the measuring facility. Alternatively, the neutron generator may be surrounded by a moderating chamber that is provided separately from the sample chamber.

Neutron flux should preferably be understood to mean the product of neutron density (free neutrons per cm 3) and neutron velocity mean (cm/s).

Neutron energy spectrum is preferably understood to mean the relative distribution of neutron energy over the entire energy range of the neutrons.

A partition should preferably be understood as meaning a predefinable/predefined spatial region within the sample, wherein the sum of all partitions yields the entire sample or defines the entire sample body. The preferred number of partitions may be selected based on the size of the sample and measurement issues, for example between 1 and 60 partitions. The volume of the partial region can be in the range of a few cubic centimeters to a few liters. In the case of very small samples, e.g. a few cubic centimeters, it may be advantageous to define only a single partition.

Photopeak efficiency should preferably be understood to mean the probability of detection of a full energy deposit for gamma emission in the detector.

Herein, gamma emission is understood to mean gamma radiation independent of its energy level. A particular gamma radiation has a particular energy. Gamma emission itself is a reaction after irradiation with neutrons. Thus, the analysis is performed from the energy spectrum of the gamma emission, in particular with respect to various types of gamma radiation. The signals of the prompt gamma radiation and the delayed gamma radiation are detected by the energy spectrum of the gamma emission.

A sample is preferably understood to mean a certain quantity of solid or liquid material selected for analysis and representative of the object under examination, including for example soil samples, ashes, water samples, sludges and chemically toxic or radioactive wastes.

The sample chamber is preferably understood to mean a component of the measuring device in which the sample is arranged during irradiation and in which the sample can optionally also be displaced, in particular during irradiation.

A sample carrier is preferably understood to mean a component of a measuring facility which receives a sample and is arranged in a sample chamber. The spatial displacement of the sample may be performed by a sample carrier.

In the following, the method according to the invention is first generally described, after which the details of the various aspects of the invention are discussed.

Operation of the one or more neutron generators causes the sample within the measurement facility to be continuously irradiated with neutrons and, simultaneously with the irradiation, gamma radiation induced/emitted due to neutron interactions is measured.

As already mentioned, gamma radiation is firstly prompt gamma radiation emitted immediately after a neutron interacts with the nuclei of the sample, and secondly delayed gamma radiation emitted after the decay of the activated nuclei according to a characteristic half-life. In one or more detectors, gamma radiation emitted by the sample can be measured in an energy-resolved manner. This will produce a measured gamma spectrum for each detector. The gamma spectrum is a record of the number of events detected in the gamma detector as a function of energy. Elements in the sample are identified based on the energy of the gamma radiation. The elemental mass can be quantified by means of the measured energy-dependent radiation intensity.

Element mass calculation in case of zoned and non-zoned measurements

After the background signal has been subtracted, the mass fraction of an element contained in the sample is calculated from the area of the light peak in the gamma spectrum caused by that element. The net photopeak count rate recorded during the multielement analysis depends on the influence parameters listed below; this relation is considered in particular in the following publications: l. Molnar (eds.), Handbook of Prompt Gamma Activation Analysis with neutron beam [ Handbook of Prompt Gamma Activation Analysis with neutron beams ], Kluyveromyces Academic Press (Kluwer Academic Publishers), ISBN 1-4020-1304-3 (2004).

(1)

Wherein

-Is the photopeak of the element with respect to the gamma energy E_γThe net count rate of (a) is,

m is the molar mass of the corresponding element,

-N_A＝6.0221408·10²³is an avogalois constant value of a constant,

-xis the position in the sample volume V,

-E_nis the energy of a neutron or a neutron,

-m(x) Is a distribution function of the mass of the corresponding element in the sample volume,

-is the position of an elementxCorresponding energy E emitted_γThe peak efficiency of the light of the gamma ray,

-σ_γ(E_n) Is a partial gamma generating cross section, and

-Φ(x,E_n) Is the neutron flux as a function of position and energy in the sample.

The partial gamma generation cross section is dependent on the element under consideration and contains both the gamma line E under consideration_γWhich in turn contains the natural frequency of the relevant isotope of the element. Since the elements in the sample to be irradiated generally emit gamma radiation of different energies, the mass is determined and used in relation toAll evaluable gamma energies E of the elements are taken into account in the analytical evaluation of the uncertainty analysis of the quality determination_γ. Considering the multiple gamma energies from one element reduces the uncertainty of the measurement method.

Preferably, the sample is subdivided into a plurality of partitions and the measurement/analysis is performed in a segmented manner. To this end, one or more sections of the sample are located in the collimated field of view of the detector during each gamma spectrum measurement (fig. 4). The field of view of the detector is the region of space with increased probability of gamma radiation detection due to the collimator geometry. In order that the respective section can be oriented in an optimal manner with respect to the field of view of the detector, the sample can be moved, in particular rotated and displaced, in front of the detector. The energy-dependent probability of detection of gamma radiation emitted from the sample or from the respective partition is called the photopeak efficiency. The gamma radiation is attenuated within the zone of the sample, so that the zone facing the detector and lying within the field of view of the collimator has a higher photopeak efficiency than the zone outside the field of view of the detector. Thus, the robustness of the measurement results with respect to the elemental composition of the respective partition can be improved, in particular if a plurality of gamma energy spectrum measurements are taken into account.

As a result of targeting up to one partition, in particular, the SNR of the corresponding partition can be improved. If the gamma spectrum measurements of multiple partitions are evaluated in combination, the uncertainty that may exist may be reduced.

It was found that a particular advantage over a total of four partitions could be produced depending on the size of the sample body. Here, the geometry of the partitions is preferably defined according to the geometry of the sample body and the measurement problem.

In a cylindrical sample, in particular, the sample is divided (fig. 4) into partitions according to layers and sectors. In particular, the partitions are produced in the form of slices of cake, i.e. as three-dimensional cylindrical sections. The horizontal sections (in particular in terms of cuts orthogonal to the axis of symmetry of the cylindrical sample) are called layers and the angularly related sections are called (angular) sectors. In an additional subdivision of the angular sectors according to the distance from the axis of symmetry, these partitions are particularly called radial sectors. In particular, a cubic or rectangular parallelepiped sample may be subdivided into individual voxels. Each voxel represents a partition. The corresponding voxels likewise have a cubic or cuboid geometry.

The quality of the elements in each partition is determined as follows. This is based in particular on the following assumptions: mass mk (symbol m) of element_k) Evenly distributed within the partition. Thus, N gamma spectra are recorded with N bins, and N net photopeak count rates are generated for each gamma energy. Depending on the type of partition, variations may occur, especially in the case of radial sectors. Then, N may be replaced by K, for example, the following applies: in the case of K bins, N gamma spectra are recorded in N measurements, where K is greater than or equal to N, and N net photopeak count rates are generated for each gamma energy. For K or N bins, in the case of a bin's collimation measurement i, equation (1) can now be simplified to the following gamma energy sum, with the index K traversing the bins, thus ensuring a simple and robust analysis:

(2)

wherein

-Is to measure the integrated peak efficiency of partition k in i,

-is the integral (n, gamma) cross section of the gamma energy in bin k during measurement i, and is formed by

And wherein

-Is the integrated neutron flux in partition k during measurement i,is formed by the following formula

Thus, a system of linear equations with dimensions N × N or N × K is generated from equation (2), which can be solved for the element masses of the individual partitions. The system of linear equations is of the form:

(3)A·m＝b

where A is a matrix of dimension NxN or NxK and m, b are vectors of dimension Nx 1 or Kx 1. The entries in matrix a are given by:

the term of m is given by mi, and the term of b is given by

In particular, in order to obtain only physically meaningful, positive and unique results for the solution of the system of equations, a so-called "non-negative least squares" method may be applied in the numerical solution of the system of equations.

Segmentation/zoning measurements may not be necessary if parameters, in particular based on the size and composition of the sample, can make simplified assumptions about the application of uniformity in mass distribution, neutron flux distribution and/or photopeak efficiency throughout the sample. It can be seen that in this case the sample can be measured in a single, preferably collimated, measurement and only a single gamma spectrum can be recorded. Equation (2) is then reduced to the following simple linear relationship:

(4)

wherein

-Is the integrated photopeak efficiency of the sample,

-is the integral (n, gamma) cross section of the gamma energy in the sample,

- Φ is the integrated neutron flux in the sample.

The equation can be solved directly for the element mass m. The corresponding parameters can be calculated in the same way for both the segmented/partitioned case and the non-segmented case.

The mass m of the corresponding element is calculated for each gamma energy, whether in a bin or in the entire sample volume. The measurement uncertainty u (m) of this value is determined in accordance with DIN ISO 11929 and can also be derived from the following publications published by the german standardization association: determination of characteristic limits for measurements of ionizing radiation (decision thresholds, detection limits and limits of confidence intervals) -rationale and application [ Determination of the characteristics limits (Determination of the detection limits and limits of the confidence intervals) for measurement of irradiation-parameters and application ] (ISO 11929:2010) (2011).

Since the elements emit gamma radiation at different gamma energies, the mass of the element is determined as a weighted average of the respective determined masses as a measurement result. The weighting is performed based on the calculated uncertainty.

Automatic identification of elements contained in a sample

Elements contained in the sample may be automatically identified based on the recorded gamma spectrum. A characteristic emission signature with a corresponding intensity of gamma energy may be created for each element from the nuclear physics database. The signal at the known gamma energy of the elements in the spectrum is compared to this signature by a computer program. Statistical analysis of the degree of correspondence provides a list of elements most likely to be included in the sample. Not only the gamma energy of the prompt gamma radiation but also the gamma energy of the delayed gamma radiation is used for this identification. In particular, the advantage is thereby created that the method can be applied to a wide variety of elements.

Method for determining photopeak efficiency

The energy-dependent photopeak efficiency of an element is determined assuming a uniform distribution of elements and mass in a section of the sample or throughout the sample. The average density of the partitions of the sample can be determined by dividing the mass of the sample by the sample volume and/or by transmission measurements by means of a gamma emitter (e.g., Co-60 or Eu-154). The transmission measurement may be used as an extended measurement for characterizing the sample, for example in order to be able to determine the filling level of the fluid to be examined or the filling in the cylinder (sample). To calculate the photopeak efficiency, a computer program is used to generate random points in the sample volume and in the next step a number of random trajectories from the points towards the detector. Then, path lengths in different materials are determined along the respective trajectories, and energy dependent probabilities of absorption and scattering along the trajectories are calculated from the attenuation coefficients of the materials. The energy-dependent probability of complete deposition of gamma energy in the detector is calculated from the path length in the detector and the energy-dependent cross-sections for the photo-electric reaction, pair-wise generation and compton scattering. The two probabilities are combined together to obtain the probability of detecting a photon along the trajectory for each possible gamma energy. The energy-dependent photopeak efficiency of this initial point comes from the average of all traces with the same initial point. The results of all energy-dependent photopeak efficiencies for an initial point in a partition are averaged to obtain a respective integrated photopeak efficiency for that partition for each gamma energy.

Method for determining neutron flux and neutron energy spectrum

For analytical evaluation for determining the elemental mass of a sample, in addition to the energy-dependent photopeak efficiency of the gamma emission, the neutron flux and the energy-resolved neutron spectrum within the sample or within the various sections of the sample are also determined by analytical methods. Here, the diffusion approximation of the linear boltzmann equation can be solved numerically. The input parameters of this set of equations are calculated from simulation calculations of neutron flux and neutron energy spectrum in an empty sample chamber and/or from neutron flux outside the sample captured by measurement.

The next paragraph describes the optional determination of the neutron flux inside and outside the sample chamber by measurement.

To evaluate the measurement data, the total or absolute neutron flux in each zone may also optionally be determined by one or more neutron detectors, which may be attached outside the sample within the measurement chamber. The total neutron flux in each individual partition can be reconstructed from the measurement data of the corresponding neutron detector. In particular, with BF3 (BF)₃) Or 3 He: (³He) gas-filled proportional counter tubes can be used as neutron detectors, in particular as neutron-sensitive materials which are particularly suitable for measuring thermal neutron fluxes. The probe may have a cylindrical geometry, in particular a length of use corresponding to the height of the measuring level. This simplifies the detection or measurement of neutron flux over the entire horizontal height. The neutron detectors are preferably arranged at the points of the expected maximum and minimum thermal neutron flux within the measurement chamber, and optionally also at additional points, preferably adjacent to the aperture of a/the collimator, in particular at the same distance from the source point (neutron source) in each case. In particular, at least four neutron detectors are arranged in the measurement chamber at the same level as the detectors. To determine the vertical distribution of the neutron flux, the same neutron detector arrangement may be provided at levels above and below the measurement level in the sense of a redundant neutron detector level. The total neutron flux in the sub-areas can be reconstructed or determined from the measurement data thus obtained, in particular taking into account the set neutron source intensity, the weight and volume of the sample and the (spatially resolved) material composition of the sample. Here, the material composition of the sample may be utilized. The reconstruction may be performed as part of an iterative evaluation process of the measurement data. In particular, a spatially resolved reconstruction is carried out taking into account the known attenuation behavior of the sample material in response to the neutron flux (in particular taking into account the attenuation coefficient) and on the basis of the characteristics of the neutron flux determined by simulation at the individual measurement points from the previously described influencing or input parameters.

Thus, in addition to gamma ray detection, neutron detection may also be performed within the sample chamber; in particular, the neutron detection is used to measure the total neutron flux in an integrated manner over an energy range, so that the respective neutron flux in the respective section can be reconstructed. The corresponding element mass can be quantified from the total neutron flux and from the neutron spectrum and the measured gamma spectrum. This provides good robustness or significance of the measurement results. Alternatively, the source intensity of the neutron generator can also be used as an input variable (in particular for the purpose of measuring redundancy) alternatively or additionally.

Typically, the neutron flux in a sample is determined in advance by determining an energy-dependent correction factor for the entire sample in an integral manner. This way of calculation is considered in particular in the following publications: trkov, g.zerovnik, l.snoj, and m.ravnik: nuclear instruments and Methods in Physics Research (Nuclear instruments and Methods in Physics Research) A, 610, page 553-565 (2009) for self-shielding coefficients in neutron activation analysis. For large volume samples, r.overtwater in a 1994 publication proposes an approximation method for determining neutron flux without energy dependence, particularly for a specific geometry that allows for reduction to two spatial dimensions, which is explained in more detail below.

The neutron flux and neutron energy spectrum within the sample or within the various sections of the sample can be determined in an automated manner (by a computer program) according to the method described herein, wherein the diffusion approximation of the linear boltzmann equation can be solved numerically in a spatially and energy-resolved manner (wherein all three spatial dimensions are taken into account), and the input parameters of this set of equations can be calculated from simulated calculations of the neutron flux and neutron energy spectrum in the empty sample chamber and/or from the neutron flux outside the sample (measured material) detected by measurement.

The spatially and energy resolved neutron flux in the sample can be determined from the solution of the complete boltzmann transport equation for neutrons:

(5)

where Ω represents a direction variable, so Ψ satisfies the following equation:

(6)

here, ∑_tRepresents the total cross section, and_swhich represent the scattering cross-sections of neutrons of the sample material, these cross-sections being made up of individual cross-sections of the sample containing elements.

Equation (6) is approximated by a system of coupled diffusion equations, corresponding to the so-called SP3 approximation; the relationship is specifically considered in the following publications: brantley and e.w.larson, Simplified P3Approximation [ The Simplified P3 application ], Nuclear Science and Engineering (Nuclear Science and Engineering), 134, pages 1-21 (2000). The set of programs is numerically solved in a plurality of sets by a computer program. As a result, the integrated neutron flux and neutron energy spectrum in each partition of the sample can be obtained, resolved in the corresponding energy bin. For determining the input parameters, an energy-resolved flux is calculated across the boundary of the sample or the measuring chamber, based on the result of a simulated calculation of the neutron flux in the empty measuring chamber. The parameters of the system of equations are derived from the elemental composition of the sample, wherein in a first step, the input parameters are calculated from the simulated neutron flux and the neutron energy spectrum in the empty sample chamber and/or from the neutron flux outside the sample detected by the measurement, in particular under the assumption about the homogeneity of the elemental composition of the sample and the sample or the zone. The neutron flux and the neutron spectrum can be calculated and evaluated separately for the respective sub-area in each case, in particular by defining the respective sub-area on the basis of a virtual subdivision of the sample into spatial regions.

Automatic iterative approach

The result of the analytical evaluation, or most importantly, is the elemental composition of the sample. The evaluation is preferably based on parameters related to the energy-dependent photopeak efficiency, neutron flux and neutron energy spectrum within the sample or within a section of the sample. Thus, a very comprehensive analysis can be performed taking into account the three calculated parameters of energy-dependent photopeak efficiency, neutron flux and neutron energy spectrum within the sample and within the sample's zones, based on the three input parameters of form/geometry and mass of the sample body and neutron source intensity. These parameters are influenced by the elemental composition of the sample, and therefore the method is preferably performed iteratively until the calculated elemental composition no longer changes, or no longer substantially changes. A possible iteration mode is described in more detail with reference to fig. 3. An initial neutron flux and an initial neutron spectrum are calculated from the input parameters (step S1). Further, the exact type of collimation and optional sample division may additionally be considered/determined, and further, translation and/or rotation of the sample in favor of the measurement procedure may also be considered (step S2). The type of collimation and sample partitioning may depend in particular on the size and geometry of the sample. Larger samples and more complex geometries are subdivided into more partitions than small samples. Evaluation of the recorded gamma spectra includes identifying the elements contained in the sample by assigning peaks measured in the spectra to individual elements, while taking into account interference between gamma peaks. The net count rate in each gamma energy is preferably calculated only once and remains constant in the iterative method (step S3). From this, the initial elemental composition of the sample was calculated. The elemental composition of the sample is determined by applying the steps to be described below for calculating the photopeak efficiency (step S4), the elemental mass in each partition of the sample (step S5), and the neutron flux and neutron spectrum (step S6). This process is performed iteratively until the elemental composition in each partition is no longer changing. As a result of the iterative analysis procedure, a high accuracy is obtained.

In particular, the net count rate of each gamma energy is calculated only once and remains constant during the iterative method. Preliminary assumptions about the elemental composition of a sample were made at the outset. The method described in the previous paragraph is then applied to determine the new elemental composition of the partitions in order to calculate the photopeak efficiency, neutron flux and neutron energy spectrum in each partition of the sample, and the elemental mass. This process may be performed iteratively until the elemental composition in each partition is no longer changing.

As a result of such an analysis procedure, the non-destructive method described here for the neutron activation-based multi-element analysis can be carried out for the first time in an automated or fully automated manner, in particular even over a relatively long duration, in particular already for large volumes of sample, in particular already exceeding about 1 liter. The form and quality of the sample body need only be taken as input parameters, taking the neutron source intensity into account in the calculation process. Here, the strength of the neutron source is determined based on operating parameters of the neutron generator or based on activity of the neutron source. Likewise, for the elemental analysis of a large volume of a sample, the respective sample is irradiated successively with neutrons, and the gamma radiation emitted by the sample and the amount of the element contained in the sample are measured, and after subtracting the background signal, in particular in a count rate-energy diagram, the element is evaluated from the area of the photopeak caused by this element. In contrast to previous methods, irradiation can be performed continuously and measurements can be performed simultaneously without pulsed irradiation or interruption of the measurement time. In addition, it is no longer necessary to know the composition of the sample for at least a portion of the sample to determine the neutron flux at the sample location. Especially for large volume samples, the sample can also be characterized by transmission measurements.

Fast neutrons are understood here to mean neutrons which, when they are emitted by the source, are fast and are subsequently decelerated by the sample chamber and/or the moderating chamber in order to be able to increase, in particular, the probability of interaction with the sample. Neutrons moderated by the moderator material may be referred to as thermalized neutrons. Thermalized neutrons are slow or moderated free neutrons, in particular with kinetic energies less than 100meV (millivolts). In the classification of neutrons, thermalized neutrons or thermal neutrons are located between cold neutrons and fast neutrons. The term "thermalization" means that neutrons reach equilibrium with the thermal motion of the medium due to repeated scattering in the medium. The velocities of these thermalized neutrons are here assumed to have a corresponding maxwell distribution, which can be described in terms of temperature.

According to one embodiment, the irradiating or irradiating and measuring is performed over a time period of at least one millisecond or at least one second. According to one embodiment, the measurement of the radiation emitted in response to the irradiation is performed in a time window of less than 5 microseconds (μ s) during the irradiation.

According to one embodiment, the irradiation is carried out without interruption for a period of at least 10 minutes or at least 30 minutes or at least two hours. Depending on the respective specific analytical problem, an optimum irradiation duration is defined. The continuous irradiation during such a time period facilitates a reliable evaluation of the measurement data in a flexible manner, in particular also in a targeted manner with respect to the individual part aspects. In particular, the irradiation duration may be defined according to the required sensitivity of the analytical problem. It has been found that the probability of detecting trace contaminants (trace contaminants) increases with the duration of the irradiation. The sensitivity of the measurement method according to the invention may increase with increasing time. In this case, the irradiation duration and the iteration may be defined independently of each other.

An irradiation duration of the order of seconds may be defined as a minimum irradiation duration. Irradiation duration in the order of seconds, minutes, hours or even days may be defined as the maximum irradiation duration.

According to one embodiment, the neutrons produced have a neutron energy value of 2.45MeV or at least one neutron energy value selected from the group consisting of: 2.45MeV, 14.1 MeV. It has been found that a particularly good signal, i.e. a signal with a favourable SNR, can be obtained in the case of neutrons of 2.45 MeV.

According to one embodiment, the neutrons generated have a neutron energy of at least one value in an energy range of 10keV to 20MeV, in particular 10keV to 10 MeV.

According to one embodiment, the neutrons produced have a neutron energy of no more than 10 MeV. In particular, this provides a high sensitivity. It has been found that the advantage of continuous irradiation with neutron energies of less than or equal to 10MeV, in particular with neutron energies of 2.45MeV, is that in this case many inelastic interactions do not occur as a threshold reaction requiring neutron energies of at least 3MeV to 4 MeV. Such inelastic interactions can lead to SNR degradation. This inelastic interaction below the chosen neutron energy can be avoided since the energy of the neutron source is now kept as low as possible. According to one possible variant of the method, neutron energies of 2.45MeV are used exclusively.

According to the invention, at least prompt gamma radiation or both prompt and delayed gamma radiation from the continuous neutron irradiation is measured and evaluated for the purpose of determining at least one element. Evaluating both types of gamma radiation expands the analysis options and increases the flexibility of the method. It has been found that it is convenient to evaluate the following reactions of irradiated nuclei without considering the time dependence of any neutron pulse: once a neutron is captured by a nucleus, gamma radiation of different energies is automatically emitted. The nuclei are de-excited by emitting a cascade of gamma emissions. The gamma radiation thus generated is characteristic of the corresponding element.

According to one embodiment, only delayed gamma radiation is measured and evaluated, at least intermittently, in response to continuous neutron irradiation, in order to determine the at least one element. This allows focusing the emphasis of the examination on certain aspects, for example in the case of samples with specific materials.

According to an embodiment, for determining the at least one element, the gamma radiation emitted by the sample is measured in an energy-resolved manner, in particular by determining a photopeak count rate, wherein such a determination comprises an energy-resolved evaluation of the measured gamma radiation from at least one gamma energy spectrum, in particular from the gamma energy spectrum detected by the respective detector. Here, the same detector can be used for both prompt and delayed gamma radiation. Energy-resolved analysis helps to improve flexibility and robustness. This also allows parallel analysis of almost all elements using one measurement.

The measured gamma spectrum is specific to the detector. The respective detectors can have a specific resolution, in each case for a specific gamma energy spectrum.

According to one embodiment, such measuring/evaluating comprises an energy-resolved measurement/evaluation of the intensity of gamma radiation emitted by the sample. This provides high flexibility and robustness, especially in combination with the evaluation of multiple types of gamma radiation.

According to one embodiment, the determining comprises an evaluation of the measured gamma radiation, wherein the evaluation comprises: at least one photopeak in the count rate-energy plot is associated with an element of the sample based on its energy. This provides a comprehensive analysis looking at both prompt and delayed radiation, in particular the gamma energy spectrum measured by one of the detectors in each case.

Here, the detected photopeak may be a photopeak characterizing prompt gamma radiation or delayed gamma radiation. During the evaluation, in particular also in the case of light peaks in which the two gamma energies interfere with one another, the prompt gamma radiation and the delayed gamma radiation can be distinguished on the basis of the respective energies. The discrimination of prompt and delayed peaks remains independent of such disturbances. For both types of gamma radiation, the respective photopeak efficiencies can be determined in this case in a spatially and energy-resolved manner by numerical methods, wherein the interaction is mapped from the emission position (source point) in the sample to the absorption in the detector.

According to one embodiment, such evaluation includes: after subtracting the background signal from the net area of a/the light peak caused by the element in the count rate-energy diagram, the mass fraction of the at least one element in the sample is quantified, in particular by means of the component of the at least one element contained in the sample being evaluated.

For this purpose, light peaks can be fitted in the energy spectrum. The area under the light peak may be defined as the background/background signal and the net light peak area may be defined as the usage signal.

According to one embodiment, the measurement of the sample, in particular of individual sections of the sample, is carried out in a collimated manner, in particular by means of at least one detector or by means of a plurality of detectors having a field of view that is specifically collimated with respect to the geometry of the sections. This may improve accuracy and also help to place emphasis on parts of the sample, especially if the SNR is good. In the case of two or more detectors, there are advantages, in particular in terms of measurement time. It has been found that the measurement time in the method described herein can be chosen shorter, and/or the sensitivity of the measurement can be increased for the same measurement time, in case more detectors are provided.

According to the invention, the sample is subdivided into a plurality of partitions, and the emitted gamma radiation is measured and evaluated with respect to the respective partitions using collimators. According to one embodiment, such measuring, determining and/or evaluating is carried out individually with respect to individual sections of the sample, which sections are manually or automatically predefined, in particular by collimation, or which can be manually or automatically predefined by collimation. This helps to focus on various areas of the sample or to simplify the evaluation of large volumes of sample or samples that are heterogeneous in composition.

Such a partition simplifies the analysis, especially in terms of the desired accuracy of the assessment. As a result, such partitions may contribute to spatially resolved elemental composition in the respective partitions. Partitioning also provides the advantage that assumptions can be made more easily or with less error. For example, in the case of a cylindrical sample body, eight or more, in particular 12, partitions are formed in each case as a cylinder section (cake slice). The detector unit then comprises, for example, two detectors which are not arranged opposite one another but are offset from one another by an angle (a circumferential angle of less than 180 °, for example a circumferential angle of 130 ° to 150 °). The entire sample can then be analyzed over the entire circumference by incremental rotations, in particular six-step rotations in 60 ° steps, for example when two detectors are used and 12 divisions are defined.

It has been found that the relation or dependency between collimation and partitioning can be used in this case. In particular, the collimation may be implemented according to the selected partition. The control means of the device may be configured to predetermine the collimation in dependence on the selected partition. Here, the following relationship may be predetermined between collimation and division: the entire partition is preferably located in the collimated field of view of the detector. The collimated field of view of the detector has only the smallest possible spatial component of the other sub-regions, i.e. the sub-regions not equal to the target position on which the collimation is focused. Due to the optionally adjustable geometry of the collimator, the field of view can be limited mainly to the target position.

Collimation has been found to be particularly effective in attenuating the background signal. Collimation measurement is understood here to mean in particular the detection of gamma radiation by at least one detector with a collimated field of view. It has been found that, also thanks to the collimation, the method described herein can be performed continuously over a considerable period of time with a particularly advantageous SNR.

According to one embodiment, such a determination comprises an evaluation of the measured gamma radiation, wherein such an evaluation is performed on the assumption that the mass and/or the element distribution in the sample is uniform, in particular in at least one of the plurality of sections of the sample. This provides a robust method, especially in the case of highly automated iterative methods.

The element and quality distributions may be assumed to be uniform at least in the respective partitions, so the respective partitions may be consistently calculated/evaluated. The measurement accuracy can also be increased by the geometric selection of the partial regions in such a way that the assumption of a uniform mass distribution applies to the most likely starting range, i.e. for example, no cake slices but the upper layers lying one above the other in the height direction.

In contrast, previous methods typically used analytical approaches where the elements were point sources. With respect to the configuration of the distribution of elements, one of two initial assumptions can be made: point sources or uniform element and mass distribution in the partitions. It has been found that the assumption of a uniform element and mass distribution in combination with the method described herein yields a very robust measurement with minimal uncertainty.

According to the invention, this determination comprises an evaluation of the measured gamma radiation, wherein this evaluation comprises: in particular, a spatially and energy-resolved calculation of the neutron flux in the respective sub-region of the sample based on a diffusion approximation of the linear boltzmann equation, in particular based on the following relationships:

in combination with the above advantages, this also provides a simple type of computation, which provides advantages especially in the case of iterations.

According to one embodiment, this evaluation further comprises calculating the neutron spectrum within the sample, in particular within a respective section of the sample, in particular in a spatially and/or energy-resolved manner, in particular based on the following relation:

this provides robustness in combination with the above advantages.

According to one embodiment, the determining comprises an evaluation of the measured gamma radiation, wherein the evaluation comprises: in particular by calculating the neutron flux and the neutron energy spectrum in an approximation method, the energy-dependent photopeak efficiency and the neutron flux and the neutron energy spectrum within the sample or the respective sub-region of the sample are calculated in each case on the basis of the following relationships:

this provides the advantages described above. It has been found that energy-dependent photopeak efficiency, neutron flux within the sample or within various partitions of the sample, and energy-resolved neutron energy spectra provide a robust basis for evaluation. The input parameters can be calculated from the neutron flux and the neutron spectrum in the empty sample chamber and/or from the neutron flux outside the sample detected by the measurement.

It has been found that diffusion approximation is particularly useful for calculations based on a small number of independent variables. This may also reduce the complexity of the analysis. A very accurate alternative method may include numerically solving the complete linear boltzmann equation in a deterministic manner or by the monte carlo method. However, the computational expense of both variants is very high; especially in the case of iteration, the expected calculation time must be in the order of hours or even days. The diffusion approximation provides a simple mathematical structure, allowing a simple numerical method to be applied.

According to one embodiment, such a determination comprises an evaluation of the measured gamma radiation, wherein such an evaluation is carried out at least partially with respect to the measured photopeak areas by means of a plurality of photopeak areas generated from a plurality of gamma energies, which are analyzed in each case for at least one element in a respective section of the sample when quantifying the mass fraction of the respective element in the respective section, said analysis being based on the following relation (for N or K sections, respectively, with the index K traversing the section):

in other words, during the evaluation, a plurality of gamma energies, each from at least one element in a respective partition of the sample, may be analyzed when quantifying the mass fraction of the respective element. This provides a simple, robust, flexible method with good accuracy. A high quality of measurement/evaluation can be ensured.

Previously, the neutron flux in a sample could be determined by determining the energy-dependent correction factor for the entire sample in an integral manner. In particular for large volume samples, it has been found that for an energy-independent determination of the neutron flux, and for a specific geometry allowing a reduction to two spatial dimensions, an approximation method can also be applied. In particular, this method is based on a diffusion equation determined by only two parameters. In particular, aspects of a method may be applied, the relationship between which has been considered in detail in the following publications: overwater, Physics of Neutron Activation Analysis in instruments for Large samples [ The Physics of Big Sample Instrument Activation Analysis ], Deivo university of chef, Delv university of technology Press, ISBN 90-407-.

In contrast, according to the invention, the neutron flux and the neutron energy spectrum within the sample or individual sections of the sample can be determined by means of a computer program, wherein the diffusion approximation of the linear boltzmann equation can be solved numerically in a spatially and energy-resolved manner, in particular taking into account all three spatial dimensions. The boundary conditions of this set of equations can be calculated from simulation calculations of neutron flux in an empty sample chamber and/or from neutron flux outside the sample detected by measurement. No energy-dependent correction coefficients need or need to be defined.

The calculation and evaluation of the neutron flux and the neutron energy spectrum can in each case be carried out separately for the respective sub-regions, in particular by defining the respective sub-regions on the basis of a virtual subdivision of the sample into spatial regions.

According to one embodiment, the method is performed based on input variables of neutron source intensity, sample geometry and sample mass, in particular exclusively on said three input variables. As a result, the method can be highly automated. Thus, only three input parameters have to be predetermined. Further parameters may be determined numerically/in an automated manner. As a result, the cost of the partial user can also be minimized. Further input parameters may be provided for calculating the neutron flux and neutron energy spectrum, for example by nuclear physics data or by simulation calculations of the input parameters.

Alternatively, a monitor or calibration material having a predetermined composition may be analyzed with the sample. This may improve the measurement accuracy or robustness of the measurement, especially in the case of samples whose composition is unknown or in the case of samples for which relatively uncertain assumptions are made. However, the use of the monitor and the evaluation of gamma radiation emitted by the monitor may optionally be implemented independently of aspects of the methods and apparatus described herein. In particular, materials that are not present in the height-determining sample may be used, such as gold in the form of a very thin gold foil placed on the sample.

According to one embodiment, the method is performed in an automated manner, in particular by evaluating the measured gamma radiation on the basis of parameters determined purely numerically, in addition to the three parameters of neutron source intensity, sample geometry and sample quality during such irradiation. This provides independence and provides the option of performing iterations in a simple manner. The method becomes more robust. The neutron flux can also be detected here by a neutron detector outside the sample.

Optionally, automation can also be implemented with respect to the three parameters described above. The sample geometry may be detected independently by the camera unit and the sample mass may be detected by the weighing unit. Both the camera unit and the weighing cell components are preferably not positioned inside the device, i.e. not within the neutron field, but outside the sample chamber, in particular outside the shielding. To this end, the device may have a measurement space for specifying the sample, in which the sample can be characterized in an automated manner. The neutron source intensity can be obtained directly from the neutron generator as a controlled variable. The neutron source intensity is directly dependent on the high voltage and current of the neutron generator. As a result of this extension of automation, a very self-contained and user-friendly device or measuring facility can be provided.

According to one embodiment, at least one measurement of the following set of measurements is performed to characterize the sample: transmission measurement, sample weighing, optical detection of the sample geometry. This firstly simplifies the operation of the measuring method for the user and secondly also simplifies the subsequent measurement, in particular in respect of the partitioning.

According to one embodiment, in particular, the method is performed iteratively in each case with respect to individual elements or with respect to the complete composition of the sample or with respect to individual partitions of the sample and/or with respect to the complete composition of the sample. This provides good accuracy, especially in the case of an easy to operate method. This also provides a method with a high degree of independence.

According to one embodiment, the spatially and energy-resolved determination of the neutron flux, in particular the absolute neutron flux, of the respective sub-region is carried out outside of the sample in the sample chamber, in particular by means of at least one neutron detector arranged in the sample chamber. This also simplifies the determination of the absolute neutron flux or the total neutron flux, whether it be an addition or an alternative to the determination made with respect to the respective sub-zone.

The invention also relates to a method for neutron activation based multi-element analysis, comprising the steps of: generating fast neutrons with energy in the range of 10keV to 10 MeV; irradiating the sample with these neutrons;

measuring gamma radiation emitted by the irradiated sample to determine at least one element in the sample; wherein the sample is continuously irradiated in a non-pulsed manner, wherein the measurement is carried out during the irradiation independently of the irradiation time, in particular simultaneously with the irradiation, wherein, for determining the at least one element, at least prompt gamma radiation or both prompt and delayed gamma radiation from the continuous neutron irradiation is measured and evaluated, wherein the evaluation is carried out on the assumption that the mass and/or the element distribution within the sample or within at least one of the plurality of sections of the sample is homogeneous. As a result, many of the above advantages result. Such measurement/evaluation may be performed independently of the time profile of the irradiation or independently of the individual phases of the irradiation.

At least one of the above objects is also achieved by a use of a detector unit comprising at least one detector in a multi-element analysis of a sample based on neutron activation, the use being configured to continuously measure both prompt and delayed gamma radiation emitted as a result of a continuous irradiation of the sample by neutrons, wherein such gamma radiation is also measured at least partially continuously, i.e. independently of irradiation time and independently of possible neutron pulses, in particular without a time window, and simultaneously with such a continuous irradiation, wherein the field of view of the detector unit is limited to a respective partition of the sample by at least one collimator, in particular is used with a plurality of detectors which are in each case collimated or sectionally collimated or adjustably collimated with respect to at least one predefinable geometry of at least one partition or with respect to at least one predefinable geometry of a partition, preferably with a collimator made of lead or bismuth. As a result, the above-described advantages are produced.

The invention also relates to the use of at least one neutron source for the multi-element analysis of a sample based on neutron activation, for generating fast neutrons for the continuous irradiation of the sample with first neutrons, these first neutrons having at least one neutron energy value from the group: 2.45MeV, 14.1 MeV; and/or irradiating the sample successively with second neutrons having a neutron energy of at least one value in the energy range from 10keV to 20MeV, in particular from 10keV to 10 MeV; and/or continuously irradiating the sample with third neutrons having a neutron energy of no more than 10 MeV. As a result, the above-described advantages are produced. Preferably, the emitted gamma radiation is detected by a detector having a collimator made of lead or bismuth.

At least one of the above objects is also achieved by comprising a control device configured to activate at least one neutron generator of an apparatus for multi-element analysis based on neutron activation, in particular an apparatus as described herein, wherein the neutron generator is configured to generate fast neutrons having an energy in the range of 10keV to 20MeV, in particular 10keV to 10MeV, wherein the control device is configured to activate the neutron generator to generate neutrons and to irradiate a sample in a non-pulsed continuous manner, in particular during at least one first time window, and wherein the control device is further configured to activate at least one detector to measure a single sample or a single piece of the sample simultaneously with such irradiation, in particular during at least one second time window independent of the first time window, continuously simultaneously with continuous irradiation and/or independently of continuous irradiation time continuously and/or intermittently Gamma radiation emitted by the sub-zones. The first time window and the second time window may be different or may be set or predefined independently of each other. The control means is further configured to limit the field of view of the detector to the respective partition of the sample by means of at least one collimator. In particular, the control device is configured to control the above-described method.

This type of analysis provides the advantages described above. The control device can synchronize at least the irradiation with the measurement and optionally also the positioning of the sample (in particular by activating/adjusting the rotation/lifting device), so that it can control the actual measurement method of the measurement facility, in particular the manner of data acquisition and the time periods thereof. The irradiation and the measurement can be carried out by means of the control device, in particular over a period of time of, for example, at least 20s or 50s, or over a number of hours or days. The control device may be coupled to the rotation/lifting device and may further be configured to position the sample arranged on the sample carrier by means of the rotation/lifting device, in particular according to or depending on the geometry of the section of the sample. Of particular importance, this provides the possibility of using a single control device to activate the entire apparatus or to allow the entire method to operate (three functional control devices), including the activation of the neutron generator, the activation of the at least one detector, and the positioning of the sample. Here, parameters such as neutron source intensity and the like may be predetermined for the neutron generator, and position data or a shift stroke and a shift speed may be predetermined for the rotating/lifting device.

The control device may be configured to control operation of a neutron generator configured to fuse deuterons for the purpose of generating fast neutrons for multi-element analysis of a sample by continuous non-pulsed irradiation of the sample.

The device or control means may comprise an input screen (user interface) or input unit for manual input of three parameters: neutron source intensity during irradiation, sample geometry, and sample mass. These parameters may also be stored in a data memory and may be read by the control means and transferred to the computer program product.

At least one of the above objects is also achieved by a computer program product for multi-element analysis based on neutron activation and configured to determine at least one element in a sample irradiated by neutrons in a non-pulsed continuous manner by evaluating gamma radiation, in particular prompt gamma radiation and/or delayed gamma radiation, emitted by the sample based on an energy-dependent photopeak efficiency and a neutron flux and neutron energy spectrum within the sample or within a single partition of the sample, and further configured to evaluate the measured gamma radiation in a partition-collimated manner by a plurality of gamma energies of the at least one element, respectively in a corresponding partition of the sample when based on a net photopeak count rate recorded during such multi-element analysis, The analysis is performed in particular when the quality scores of the respective elements of the respective partitions are quantified on the basis of the following relations:

the set of equations relates here to the determination of the quality of the elements in the individual partitions according to the invention (evaluation of the equations for all qualities of the individual partitions, in particular simultaneously). The computer program product facilitates a high degree of automation or independence in combination with a high degree of analytical accuracy. In particular, the computer program product is configured to perform the above-mentioned types of evaluation in an automated manner. In this context, the evaluation of the particle alignment is understood to mean that the individual regions of the sample, which have been defined in advance in terms of geometry and are delimited from one another, are evaluated individually.

In addition, the computer program product may also be configured to predetermine the desired position of the sample, in particular depending on the detected or input sample geometry, in particular based on the desired position stored in the position database, depending on the sample geometry and/or the sample size, wherein the positioning may be carried out in particular by activating/adjusting the rotating/lifting means.

The computer program product may be configured to calculate various variants of the partitions for the respective sample geometries and propose the partition identified as optimal or to directly autonomously select such a partition. To this end, the computer program product may perform an approximate uncertainty analysis according to the number and configuration of partitions, and may set partitions according to the accuracy required by the user, or autonomously determine such partitions. In this procedure, the configurable collimator may be specifically set with respect to the selected partition.

The invention further relates to a data medium on which such a computer program product is stored, or a computer system or a virtual machine or at least one hardware element with the computer program product.

The invention also relates to a computer program configured to provide the evaluation means described herein or to provide the method steps described herein in this respect.

According to an exemplary embodiment, the computer program product is configured to evaluate the integral measurement based on a single gamma spectrum, in particular with respect to the non-partitioned sample, in particular based on the following relation:

this provides a simplification of the assumptions about uniform mass distribution.

The method according to the invention can also be described as follows. In a method for the non-destructive elemental analysis of a sample, the respective sample is continuously irradiated with fast neutrons, wherein the gamma radiation emitted by the sample is measured simultaneously with the irradiation, wherein the amount of the element contained in the sample is evaluated from the net area of the photopeak after subtracting the background signal caused by the element in the count rate-energy diagram. In this case, the quantification of the elemental mass of the sample can be carried out in an automated manner, and the parameters required for the analysis, in addition to the neutron source intensity during irradiation and the geometry and total mass of the sample, can be calculated numerically. Here, no monitor for neutron flux or internal or external calibration standards of the sample is required. Here, the sample may be decomposed into spatial regions (partitions), and each partition of the sample may be measured in a collimated manner. Here, the determination of the elemental mass of the sample may be based on the assumption that the elemental distribution and the mass distribution in the respective partitions of the sample are uniform. Here, the average density of the partitions can be derived by transmission measurements using radioactive gamma emitters. The neutron flux in the sample compartment can be determined by an analytical method which numerically solves the diffusion approximation of the linear boltzmann equation and calculates the boundary conditions of this set of equations from the result of a simulated calculation of the neutron flux in the empty sample chamber and/or the neutron flux outside the sample detected by measurement. The photopeak efficiency can be determined here in a spatially and energy-resolved manner by numerical methods, wherein the interaction is mapped from the emission (source point) in the sample to the absorption in the detector. Here, the method may be iteratively performed with respect to the components of the sample until the calculated sample components stabilize. In this case, all detected light peak areas, which result from the various gamma emissions of the elements in the sample, can be taken into account in the analytical evaluation. In this case, the measurement result of each partial region can be taken into account in the analytical evaluation, as a result of which the sensitivity and accuracy of the measurement method of the entire sample can be increased. The neutron source and the sample can be located in a sample chamber made of graphite, which is embodied as a moderating chamber. In this process, an effective shield of neutron radiation may be located directly around a/the moderating chamber or around the sample chamber. The detector or the detector unit can be located in a collimator made of a material that shields gamma rays. Here, the geometry of the sample and the moderating chamber and/or the sample carrier may reduce the neutron flux gradient within the sample, which may be obtained in particular by a variable moderating length (path length between neutron source/source point and sample), wherein the influence of the neutron flux gradient is reduced or changed.

At least one of the foregoing objects is also achieved by an apparatus configured to perform neutron activation based multi-element analysis, comprising:

a neutron generator configured to generate fast neutrons;

a sample chamber and a sample holder disposed therein;

a detector unit comprising at least one detector configured to measure gamma radiation emitted by the irradiated sample in order to determine at least one element in the sample; wherein the device is configured to irradiate a certain/the sample arranged on the sample carrier in a non-pulsed continuous manner and to measure prompt gamma radiation and/or delayed gamma radiation emitted by the irradiated sample independently of the irradiation time, in particular without a time window, in particular simultaneously with a continuous irradiation during irradiation, in particular to perform the above-described method. As a result, the above-mentioned advantages arise, in particular in combination with a low background signal or a background signal which has been minimized by means of apparatus and/or method techniques.

Here, the apparatus comprises at least one collimator limiting the field of view of the detector to respective partitions of the sample, and is configured to subdivide the sample into individual partitions, and is further configured to measure at least prompt gamma radiation or both prompt and delayed gamma radiation emitted by the continuously irradiated sample with respect to the respective partitions of the sample during such irradiation. The apparatus further comprises: a control device configured to perform automatic continuous irradiation and configured to control/adjust automatic measurement of neutron irradiation continuously applied during such irradiation. The apparatus is further configured to determine neutron flux within respective partitions of the sample in a spatially and energy resolved manner and to evaluate measurements of the partitions by quantifying a mass fraction of the at least one element in the sample. This also simplifies automation.

Preferably, the semiconductor detector or the scintillation detector is used as a separate detector (of the detector unit), i.e. a detector with a high energy resolution, which is configured to measure prompt gamma radiation and delayed gamma radiation.

Optionally, the method is modified by a moderating chamber provided independently of the sample chamber. As a standard, the moderating chamber may be provided/installed in a spatially fixed apparatus. The moderating process can be performed in the sample chamber, the moderating chamber, and/or the sample itself, as desired.

The respective detectors of the device may be focused by at least one collimator. A collimator configured to predetermine or define or set the field of view of the detector provides in particular the advantage of an improved SNR, in particular also in combination with a continuous irradiation. Furthermore, the sample can be measured in a zonal alignment.

Preferably, the plurality of detectors are arranged at the same height level, in particular at the height level of the neutron source or the source point. Preferably, the detector or detectors are arranged as close as possible to the source point. This may provide good measurement results or help to minimize background signals. In the case of horizontal arrangement at the same height, the detectors are preferably offset in the circumferential direction by less than 90 °, for example by 60 ° or 75 °, from the neutron source point in plan view.

The neutron source point is preferably understood here to mean the position or orientation at which neutrons are emitted, in particular towards the sample into the sample chamber. The neutron generator may be arranged independently of the location of the neutron source point, or the location of the neutron source point may be predetermined.

Like a/the slowing chamber, the collimator may be fixedly mounted in a single predefined setting or configuration. Optionally, the collimator may also have a plurality of settings for a predeterminable field of view, respectively, for example a first setting with a relatively wide/wide field of view, a second setting with an intermediate field of view and a third setting with a relatively narrow/tight/concentrated field of view, wherein the collimator may be switchable between these settings.

Here, the device comprises at least one component attenuating a background signal of the device, said component being selected from the group consisting of: at least one collimator, preferably made of lead or bismuth, which limits the field of view of the (respective) detector to a partition of the sample. The apparatus may further comprise: a moderator chamber made of graphite, and/or a shield made of boron-containing polyethylene, and/or a sample chamber and/or a sample carrier each made at least partially of graphite or a perfluorinated plastic or beryllium. As a result, in particular, an improved SNR can be obtained. The collimator preferably has a wall strength of at least 5 cm. The device can analyze the sample in a non-destructive manner and evaluate the emitted gamma radiation from a number of aspects in the process. The device is not limited to the evaluation of a specific type of gamma radiation or to the evaluation within a certain time window.

It has been found that an advantageous angle between the neutron generator and the detector is between 50 ° and 90 °, in particular because this avoids exposing the detector to too high a neutron flux. Furthermore, detectors in this angular range can be focused in a spatial region in which the neutron flux in the sample is highest where possible.

Previously, gamma radiation resulting from inelastic interactions (scattering processes) in previously employed facilities produced strong background signals such that gamma radiation could be detected only after a certain waiting time (time window) after the neutron pulse. The previously used detectors or detectors in previously used facilities were previously effectively "blind" after the corresponding neutron pulse. Too high a signal rate results in detector failure. In each case, the detector then requires a certain waiting time before detection can be carried out again, i.e. until the signal rate drops again.

According to an exemplary embodiment, the apparatus further comprises a computer program product or a data storage with the computer program product, wherein the computer program product is configured to determine the at least one element in the sample by evaluating the measured gamma radiation based on the energy-related photopeak efficiency and the neutron flux and neutron energy spectrum within the sample or within the respective section of the sample, in particular based on at least one of the relations that have been previously described with respect to the computer program product. These formulas or relations may each be stored as one of a plurality of calculation bases in a data memory, through which the computer program product can interact, or which the computer program product can access. In addition to having a large flexibility during determination/evaluation, this also provides the option of a fully automated iterative multi-element analysis, in particular in combination with a control device for controlling/regulating the neutron emission, the detector and/or the rotation/lifting device.

According to an exemplary embodiment, the device further comprises a rotation and/or lifting means configured to displace the sample carrier or sample in a translational and/or rotational manner, preferably a rotation and/or lifting means decoupled from a/the sample chamber of the device, wherein the at least one electric drive means of the rotation/lifting means is arranged outside the shielding of the device (in particular outside the shielding made of boron-containing polyethylene). This also provides good SNR.

According to an exemplary embodiment, the device further comprises a unit for transmission measurement configured to determine an average density of the sample or the respective partition. The transmission unit comprises a radioactive gamma emitter (in particular Eu-154 or Co-60) and a detector for measuring the attenuation of gamma radiation after penetration of the sample. Here, the detector for measuring the attenuation of gamma radiation may be one of the detectors for prompt gamma radiation and delayed gamma radiation, or may be a detector provided specifically for such transmission measurement. During transmission measurements, the sample is not irradiated by neutrons. This helps to distinguish the type of gamma radiation.

According to an exemplary embodiment, the apparatus comprises at least two detectors arranged point-symmetrically, in particular with respect to the neutron generator or with respect to at least one neutron source or at least one neutron source of the apparatus. As a result, the respective sub-area can be optimally positioned or oriented in front of the respective detector. In addition, these partitions may be geometrically defined depending on the geometry of the sample, and the sample may be oriented accordingly, e.g., rotated once over its entire circumference in six rotation steps.

The translation and/or rotation may be performed in such a way that the center of the respective partition is at the same level as the detector, or along the visual axis of the detector, or that the sample or partition is in the collimated field of view of the detector.

According to an exemplary embodiment, the apparatus further comprises a control device configured to perform an automatic continuous irradiation and/or to control/adjust the automatic measurement in case of such a continuously applied neutron irradiation independently of said neutron irradiation time, simultaneously with such a continuous irradiation, in particular to perform an iterative automatic evaluation of the emitted and measured gamma radiation independently of the neutron irradiation time based on parameters (with the exception of three parameters which may be predetermined manually or automatically, i.e. neutron source intensity, sample geometry and sample quality during irradiation) which are only numerically determined or read by the control device. This provides the option of allowing individual steps or the entire method to be performed in an automated manner.

According to an exemplary embodiment, the neutron generator comprises a neutron source or source point configured to fuse deuterons (deuterium nuclei), in particular using deuterium gas as gaseous target or gaseous fuel. It has been found that a sufficiently high source strength can be ensured even by fusion of deuterons. The use of this energy range provides advantages both during continuous irradiation and during measurement and evaluation, firstly due to low neutron energy and secondly long irradiation duration. The fuel may be gaseous (rather than solid) and therefore no longer needs to replace a solid target (solid) that has run out after a certain period of time. In particular, this simplifies the analysis performed over a long period of time and can ensure high reproducibility or high reliability of the measurement results.

According to an exemplary embodiment, the neutron generator is an electrically operated neutron generator or comprises at least one radionuclide neutron source, such as an AmBe source. A neutron generator that fuses deuterons and emits neutrons with a starting energy of 2.45MeV as a result of the fusion reaction is preferred. In this case, pulsed irradiation can also optionally be carried out in particular at 2.45 MeV. In contrast, for neutrons with just this energy value, a neutron generator that fuses tritium-deuterium (14.1MeV) is often used in the previous pulse irradiation.

According to an exemplary embodiment, the at least one detector is a semiconductor detector or a scintillation detector. This facilitates accurate assessment of both prompt and delayed gamma radiation over a wide energy range.

The invention also relates to an apparatus for neutron activation based multi-element analysis, the apparatus comprising:

a neutron generator configured to generate fast neutrons;

a sample chamber and a sample holder disposed therein;

a detector unit comprising at least one detector configured to measure gamma radiation emitted by the irradiated sample in order to determine at least one element in the sample; wherein the apparatus is configured to irradiate a/the sample in a non-pulsed continuous manner and to measure prompt gamma radiation and/or delayed gamma radiation emitted by the irradiated sample independently of irradiation time during such irradiation, wherein the apparatus comprises at least one component attenuating a background signal of the apparatus, said component being selected from the group of: at least one collimator limiting the field of view of the detector to the sample or to a partition, and/or a moderator chamber made of graphite, and/or a shield made of boron-containing polyethylene, and/or a sample chamber and/or a sample carrier each made at least partially of graphite or perfluorinated plastic or beryllium, wherein the apparatus further comprises a control device configured to perform an automatic continuous irradiation and/or configured to control/adjust the automatic measurement independently of the respective phase time of the neutron irradiation, in particular simultaneously with such a continuous irradiation, during such an irradiation in case of a continuously applied neutron irradiation. As a result, many of the above advantages result.

The above advantages are achieved due to the use of a neutron generator in multi-element analysis of a sample based on neutron activation, the neutron generator being configured to fuse deuterons, in particular as a gaseous target or gaseous fuel, to generate fast neutrons, in order to subject the sample to continuous non-pulsed irradiation.

Drawings

The invention will be described in more detail in the following figures, to which reference is made for reference numerals not explicitly described in the respective figures. The details are as follows:

FIG. 1 shows a perspective view of a schematic illustration of an apparatus for nondestructive multi-element analysis according to an exemplary embodiment;

fig. 2A, 2B, 2C show in each case a cross-sectional view of a sample chamber with one or two detectors and a detailed view of the detectors of an apparatus for nondestructive multi-element analysis according to an exemplary embodiment;

fig. 3 shows a schematic illustration of the steps of a method according to an embodiment in a flow chart;

fig. 4 shows a cylindrical sample with partitions in the form of disk segments, as an example of partitioning in a method according to an embodiment;

FIG. 5 shows a cross-sectional view of a sample chamber of a device according to an exemplary embodiment, the sample chamber having a rotation and elevation means arranged outside the shield; and is

Fig. 6 shows a schematic representation of a sample cell of an apparatus for nondestructive multi-element analysis according to an exemplary embodiment, the sample cell having a neutron detector of a neutron detector unit disposed therein.

List of reference numerals

1 sample of

Apparatus for neutron activation based multi-element analysis

11 neutron generator

11.1 neutron sources or points

12 moderating chamber

14 sample carrier

15 sample chamber

16 detector unit

16A, 16B detectors

16.1 Crystal of Individual Detector

16.2 Detector end cover

16.3 Crystal holder

17 collimator

18 rotating and lifting device

18.1 coupling, in particular shaft

19 shield

20 control device

21 data memory

22-core physical database

23 input unit/screen

24 transmission measuring cell

25 Camera Unit

27 weighing cell

28 neutron detector unit

28A, 28B, 28C, 28D neutron detectors

30 computer program product

Visual axis of A16 detector

M1 Material 1, in particular boron-containing polyethylene and/or concrete

M2 Material 2, in particular lead and/or bismuth

M3 Material 3, in particular graphite

M4 Material or Medium 4, in particular air

M6 material 6, in particular lithium polyethylene and/or lithium silicone

M7 material 7, in particular aluminium and/or carbon fibre reinforced plastic

M8 Material 8, in particular copper or Plastic

M10 material 10, in particular germanium

Partitioning of P1, P2, Pn samples

R1 first control point

R2 second control Point

R3 third control Point

Fourth control Point R4

Fifth control Point R5

S1 first step, in particular producing and irradiating neutrons

S1.1 setting (controlling or regulating) neutron source intensity

S1.2 slowing down

S1.3 calculating neutron spectra by simulation

S1.4 calculating neutron flux by simulation

S2 second step, in particular sample assignment and measurement

S2.1 probing sample mass and/or sample geometry and/or Transmission measurements

S2.2 collimation

S2.3 detecting or setting sample zones

S2.4 Shifting/positioning of the sample, in particular by translation and/or rotation

S3 third step, in particular detecting/measuring emitted gamma radiation and evaluating the measured gamma radiation

S3.1 detecting/measuring Gamma radiation or evaluating Gamma Spectrum

S3.2 element/Peak identification

S3.3 interference analysis

S3.4 evaluation of the Peak, in particular in terms of Peak area and background

S4 fourth step, in particular evaluating the measured gamma radiation

S4.1 evaluation of interactions within samples

S4.2 evaluating interactions in detectors

S4.3 determining the solid Angle between the sample and the Detector

S4.4 determining the photopeak efficiency, in particular the initial photopeak efficiency

S5 fifth step, in particular determining at least one element, in particular mass

S5.1 determining at least one element mass or element mass ratio

S5.2 determining at least one section

S6 sixth step, in particular computational neutronics

S6.1 evaluation of interactions within samples

S6.2 evaluation of neutron spectra

S6.3 evaluation of neutron flux

v1 first variable/parameter, in particular capable of being manually input, in particular neutron source intensity

v2 second variable/parameter, in particular capable of being manually entered, in particular sample geometry

v3 third variable/parameter, in particular able to be manually entered, in particular sample quality

x longitudinal axis

y transverse axis

z vertical axis

Detailed Description

Fig. 1 shows components of an apparatus 10 for non-destructive multi-element analysis, in particular, components of a measurement facility for performing the method for neutron activation based multi-element analysis described herein.

By operating one or more neutron generators 11, the sample 1 is continuously irradiated with neutrons and the gamma radiation induced/emitted thereby is measured simultaneously with the irradiation. The device/measuring installation 10 comprising the sample 1 is composed in particular of the following components: the neutron generator 11 comprises at least one electrically operated neutron source, in particular a source of neutrons which fuses at least deuterium and deuterium (or deuterons) and optionally promotes the fusion of another type, in particular tritium and deuterium. Fast neutrons with energy of 2.45MeV are emitted during the deuteron fusion reaction. Here, deuterium gas is preferably used as target (non-radioactive). Optionally, at least one further energy value, in particular 14.1MeV, may be provided by means of a neutron generator. The neutron generator 11 is located within the moderating chamber 12 and is surrounded by a shield 19. The moderating chamber 12 is constructed of a material, preferably graphite, that is as effective as possible in moderating fast neutrons and emits as little gamma radiation as possible during the moderating process. Gamma radiation not emitted by the sample but still recorded by the detector is defined as an active background signal. The apparatus 10 described herein advantageously provides very weak, minimized background signals, and thus can measure gamma radiation very flexibly.

During irradiation, the sample 1 is located on the sample carrier 14 inside the sample chamber 15. The sample carrier may be, for example, a rotating plate, a cartridge, a pot, or a flask. Preferably, graphite and perfluorinated plastics may be used as the material of the sample carrier 14.

The sample carrier 14 and the sample chamber 15 are designed such that the sample is irradiated as uniformly as possible by neutrons (i.e. with a small local neutron flux gradient) and neutrons deviating from the sample are effectively reflected back into the sample. During the interaction between the neutrons and the sample carrier 14 and between the neutrons and the sample chamber 15, an active background signal should be induced which may only be weak. In particular, this can be ensured by using graphite, beryllium and also perfluorinated plastics or carbon fiber-reinforced plastics as the material of the sample carrier 14 and the sample chamber 15.

The gamma energy spectrum measured simultaneously with the irradiation is registered by the detector unit 16 or by one or more detectors 16A, 16B. Both a detector and a plurality of detectors are understood to mean a detector unit 16. The measurement time of the sample can be reduced by a plurality of detectors or the sensitivity and accuracy of the multi-element analysis method can be improved without changing the measurement time. The detector unit 16 records the energy of the gamma radiation emitted by the sample and counts the energy deposits in the detector. A collimator 17 is located around each detector 16. A corresponding collimator may be used to limit the "field of view" of the detector employed in such a way that the gamma radiation emitted by the sample is predominantly detected. The spatial region with an improved probability of detection of gamma radiation has in particular the form of a cone or pyramid starting from the detector. The collimator 17 is made of a material, preferably lead, which shields gamma radiation as effectively as possible. The collimator allows minimizing or attenuating the active background signal due to the limited field of view of the detector.

The sample may be measured in a segmented/zoned manner. For this reason, during a single gamma spectral measurement, not the entire sample body, but only the individual segments, so-called partitions, are located in the collimated field of view of the detector. To position the various zones in the field of view of the detector, a rotation and lifting device 18 is provided to rotate and/or translate the sample and sample carrier. The rotation and lifting device and the sample carrier are connected to one another in particular in a press-fit and/or interlocking manner. Since the components of the rotation and elevation device may increase the active background signal, these components are preferably positioned outside of the sample chamber 15 and the moderating chamber 12 and outside of the shield 19 (FIG. 5). In particular, a shaft, chain or toothed belt may be used to transmit forces between the rotating and lifting device and the sample carrier 14.

Shields 19 are arranged around the moderator chamber 12 and the sample chamber 15 and around the detector unit 16 and the collimator 17. The shield 19 surrounds the measurement facility and reduces the ambient neutron and gamma dose rates outside the measurement facility. Preferably, boron-containing polyethylene is used as the material of the part of the shield that is mainly shielding neutron radiation. Concrete and elements with higher atomic numbers and higher densities (e.g., steel or lead) can be used as the material of the shield that primarily reduces or attenuates gamma radiation. It has been found that in and around the region of the moderator chamber 12 and the sample chamber 15, as well as around the detector unit 16 and around the collimator 17, a significant improvement of the SNR can be achieved with boron-containing polyethylene.

Fig. 1 further shows that at least one of the at least three variables/parameters v1, v2, v3, in particular the neutron source intensity, the sample geometry and/or the sample mass, can be input or called for at the input screen 23. According to a variant, these three parameters can also be determined by the device 10 in a fully automatic manner.

In the arrangement shown in FIG. 1, the moderation may be performed in a separate moderating chamber 12 outside of sample chamber 15. Optionally, the moderation may also be performed within sample chamber 15. Generally, the moderation may be performed in the moderating chamber 12, the sample chamber 15, and/or the sample 1 itself.

Fig. 1 further shows a transmission measurement unit 24, by means of which additional characterization of the sample can optionally be carried out, in particular on the basis of gamma radiation.

Fig. 1 further shows components for automating the measurement or evaluation, in particular a control device 20, which is coupled to a data memory 21, a nuclear physics database 22, an input unit/input screen 23, a transmission measurement unit 24, a camera unit 25, a weighing unit 27, and/or a computer program product 30, wherein the computer program product may also be stored in the control device 20.

Fig. 2A, 2B, 2C show an apparatus for non-destructive multi-element analysis by which collimated measurements are made on a section of a sample. In the variant shown in fig. 2A, the two detectors 16A, 16B are arranged symmetrically with respect to the neutron source or with respect to the neutron source point 11.1 of the neutron generator 11.

In detail, fig. 2A, 2B, 2C also show the materials that can preferably be used, in particular for ensuring a weak background signal, in particular a boron-containing polyethylene M1 (in particular 5% or 10%) of the shield 19 for neutron radiation (and segmented concrete of the shield 19 for gamma radiation), lead or bismuth M2 for the collimator 17 or for the purpose of shielding gamma radiation, graphite M3 for the slowing chamber 12 or the sample carrier 14 or the sample chamber 15, lithium-6-polyethylene or lithium-6-silicone M6 for shielding the detector, germanium M10 for the crystal 16.1. The regions between the individual components of the detector 16, in particular between the detector end cap 16.2 and the crystal 16.1, in particular the interior of the collimator, are filled with air M4. Depending on the type of neutron generator or detector, suitable materials M7, M8 may be selected for the further individual components, in particular from the list comprising copper, aluminum, plastic (in particular carbon fiber reinforced plastic).

Fig. 2A shows, by way of example, two sections P1, P2 of the n sections Pn in the form of cylinder sections (cake slices) in the case of a cylindrical sample 1. The sample 1 can here also be provided by a cylinder with a filling level of a filling or fluid. The cylinder may have a relatively large volume, for example 200 litres.

Furthermore, fig. 2A allows the orientation of the various components to be identified from the longitudinal axis x, the transverse axis y and the vertical axis z. The sample is at least partially cylindrical or, for example, embodied as a cylinder, and it extends along a vertical axis z, in particular in a rotationally symmetrical manner about the z-axis. Positioning at different z levels is possible by means of the aforementioned lifting device 18.

Fig. 2B shows a variant with only one detector 16, which is collimated on a cylinder segment. Such an arrangement may in particular also be provided in a cost-optimized manner.

In the arrangements shown in fig. 2A and 2B, respectively, the moderation may also be performed exclusively within sample chamber 15.

Fig. 2C further shows a detector end cap 16.2 and a crystal holder 16.3, by means of which the crystal 16.1 can be positioned and oriented.

Fig. 3 shows a method with six steps S1 to S6, which steps each comprise sub-steps. Control points R1 through R5 may be provided between the various steps, whether for user queries or for automated computer-controlled queries.

Generating neutrons and irradiating the sample with neutrons is performed in a first step S1, wherein the first step may comprise at least one of the following sub-steps: neutron source intensity is set (controlled or adjusted) (S1.1), moderated (S1.2), neutron spectrum is calculated by simulation (S1.3), and neutron flux is calculated by simulation (S1.4). In particular, there may be an optionally repeated query regarding the neutron source intensity at the first control point R1, either an automatic data query or within the scope of user input/user guidance.

The sample is specified and measured in a second step S2, wherein the second step may comprise at least one of the following sub-steps: the sample mass is detected and optionally also the sample geometry (S2.1), collimated (S2.2), the sample partition is detected or set (S2.3), in particular the sample is displaced/positioned by translation and/or rotation (S2.4). Step S2.1 may be carried out in connection with transmission measurements, in particular by means of emitting radioactive gamma radiation towards the sample, for example for detecting the filling level in the cylinder (sample) or for determining the matrix density. Transmission measurements may also be considered extended measurements for characterizing the sample and may provide as useful additional data as possible, in particular also with respect to the subareas. In particular, there may be an optionally repeated query at the second control point R2 regarding sample mass, sample geometry and partition, either an automatic data query in communication with the camera unit and/or weighing unit, or within the scope of user input/user guidance. In particular, a positioning or orientation of the sample can also be carried out at the second control point R2.

The emitted gamma radiation is detected or measured in a third step S3, wherein the third step may comprise at least one of the following sub-steps: gamma radiation is detected/measured and the gamma spectrum is evaluated (S3.1), element/peak identification (S3.2), interference analysis (S3.3), peak evaluation, in particular with respect to area and background (S3.4). In particular, the transmission and verification of the intermediate result may be carried out at the third control point R3. Here, the control point R3 may comprise a plausibility check, in particular within the scope of statements about the homogeneous element and mass distribution in the sample or the respective section, wherein there may optionally be iterations back to step S2 (for example in the case of deviations greater than a maximum threshold), in particular in order to measure on the basis of a new alignment route.

The measured gamma radiation is evaluated in a fourth step S4, in particular in order to calculate an energy-dependent photopeak efficiency, wherein the fourth step may comprise at least one of the following sub-steps: evaluating the interaction within the sample (S4.1) in order to calculate an energy-dependent photopeak efficiency, evaluating the interaction in the respective detector (S4.2), determining the solid angle between the sample and the detector (S4.3), determining the photopeak efficiency, in particular the (initial) photopeak efficiency (S4.4). In particular, the transmission and verification of the intermediate result may be carried out at the fourth control point R4.

In a fifth step S5, the quality of at least one element is determined, wherein the fifth step may comprise at least one of the following sub-steps: determining at least one element mass or determining an element mass ratio (S5.1), determining at least one cross section (S5.2), in particular according to step S1 or step S4, respectively. In particular, the transmission and verification of the intermediate result may be carried out at the fifth control point R5. Here, the control point R5 may comprise a plausibility check, in particular a juxtaposition or comparison of the quantified elemental mass with the total mass of the sample.

Neutronics is calculated in a sixth step S6, wherein the sixth step may include at least one of the following sub-steps: the interaction, in particular the neutron interaction within the sample, is evaluated, in particular by diffusion approximation (S6.1), the neutron spectrum is evaluated (S6.2), in particular the neutron flux is evaluated by diffusion approximation (S6.3).

The control points R1 to R5 may each include optional feedback (control loops) to previous steps, particularly within the scope of user input or verification of intermediate results. Steps S4 to S6 may be performed iteratively independently of the respective control points, in particular continuously during the evaluation of gamma radiation from the continuously irradiated sample. The iteration is terminated if the element quality to be determined no longer changes or at least no longer changes substantially, for example below a predeterminable threshold value for the difference.

Fig. 4 shows the field of view of the respective detector 16A, 16B of an example using a cylindrical sample 1 which has been partitioned into circular discs and circular sections P1, P2, Pn. Here, the field of view of the respective detector 16A, 16B does not necessarily correspond to or be flush with the respective sector. The extent of neighboring regions in the field of view of the respective detector, which are intended to be evaluated simultaneously or removed by calculation, may be taken into account during the evaluation. There are 12 partitions per level. The entire sample can then be analyzed by six rotations and a corresponding number of translational horizontal displacement steps (here, there are five horizontal elevations, i.e. four displacement steps in the z-direction). For example, each partition is irradiated and measured over a period of several seconds to several minutes.

Fig. 5 shows an apparatus 10 in which the sample carrier 14 can be displaced upwards by a considerable distance in terms of its level (dashed arrow). The sample chamber 15 is defined by a material M3, which material M3 can be displaced together with the sample 1 into an air-filled cavity above the sample 1. The rotation and lifting device 18 is connected to the sample carrier 14 by a coupling comprising a shaft 18.1; however, in addition to this, the rotating and lifting device is arranged outside the neutron shield and is isolated from the sample chamber. This ensures that neutrons do not reach the rotating and lifting device. The path of the neutrons to the rotating and lifting means is blocked. The material of the guide shaft 18.1 is preferably graphite. As shown in fig. 5, a passage for the shaft 18.1 may be provided in the graphite block. Preferably, the rotation and lifting means 18 are connected to the sample carrier 14 only by the shaft 18.1. In this case, the neutron shield is penetrated only by the shaft. The rotating and lifting device is arranged behind the effective neutron shield in an isolated manner.

Fig. 6 shows an apparatus 10 for non-destructive multi-element analysis, wherein four neutron detectors 28A, 28B, 28C, 28D of a neutron detector unit 28 are arranged in a sample chamber 15. The neutron detectors, here arranged in an exemplary manner, are arranged in a uniformly distributed manner around the circumference of the sample chamber 15. Optionally, more than four neutron detectors may also be provided. The neutron detector is preferably arranged at the installation level of the gamma detector. The neutron detector unit 28 can carry out a spatially and energy-resolved determination of the neutron flux, in particular of the absolute or total neutron flux of the respective sub-zone, outside the sample.