JP2013120123A - Nuclide composition analyzer, and nuclide composition analysis method - Google Patents

Abstract

PROBLEM TO BE SOLVED: To non-contactly, nondestructively and precisely perform nuclide composition analysis in a specimen containing a mix of many nuclides.SOLUTION: An incident neutron beam 20 is irradiated to a specimen 100 (neutron irradiation process). A gamma ray detector 13 detects a neutron capture gamma ray 30 occurring from the specimen 100 at a plurality of points of time with the occurrence time of the incident neutron beam 20 as an origin (gamma ray detection process). Next, a nuclide composition generating the neutron capture gamma ray 30 is temporarily calculated by calculating the intensity of the neutron capture gamma ray at the respective points of time. Then, the nuclide composition is calculated by performing statistical treatment of the composition temporarily calculated at each point of time (first composition calculation process). When calculating a more accurate composition from the temporary composition at each point of time, fitting or the like obtained by taking an error into account can be performed.

Description

  The present invention relates to a nuclide composition analyzing apparatus and a nuclide composition analyzing method for analyzing a nuclide composition in a sample containing a plurality of types of radioactive materials in a non-destructive and non-contact manner.

  In the composition analysis of molten nuclear fuel generated by a nuclear accident or the like, the number of nuclides (types of radioactive materials) to be analyzed becomes very large. In addition, since the molten nuclear fuel is usually present in the shield under an environment with a high radiation dose, it is necessary to analyze the sample in a non-destructive manner without contacting the sample.

Non-patent Document 1 and Patent Document 1 describe a technique (first prior art) for analyzing the composition of nuclides by examining the neutron resonance absorption reaction of a nucleus as a technique capable of performing such an analysis. . In this technique, the composition analysis of the nuclide can be performed by irradiating the sample with a neutron beam and analyzing the absorption line in the spectrum of the neutron beam after passing through the sample. The spectrum of a neutron beam can be measured by, for example, the TOF (Time of Flight) method using a difference in detection time according to the energy (velocity) of neutrons. This technique is effective for nuclides having a large reaction cross section of neutron resonance absorption for neutrons having an energy of about eV, and for example, composition analysis of ²³⁵ U, ²³⁸ U, etc. can be performed with high accuracy.

Further, Patent Document 2 discloses a technique (second conventional technique) for performing composition analysis by irradiating a sample with neutron rays and detecting γ rays (neutron capture γ rays) generated by a neutron absorption reaction of a nucleus. It is described in. As the neutron beam, a neutron having a higher energy than that used in the first prior art is used. For example, an epithermal neutron obtained by reducing a fast neutron emitted from ²⁵² Cf to 10 to 100 eV is used. In this technique, ¹⁰ B, which has a particularly high neutron absorption capacity and is contained in the molten nuclear fuel, can be detected with high accuracy. ^When detecting ¹⁰ B, 478 keV γ-rays are detected as neutron capture γ-rays.

J. et al. W. Bahrens, R.A. G. Johnson, and R.D. A. Schrack, “Neutron Resonance Transmission Analysis of Reactor Fuel Samples”, Nuclear Technology, vol. 67, p162, 1984

JP 2006-10356 A JP 2001-242101 A

However, for example, molten nuclear fuel often contains both ²³⁵ U and ¹⁰ B. In such a case, for example, in the first prior art, analysis of ²³⁵ U or the like can be performed, but analysis of ¹⁰ B or the like that does not cause neutron resonance absorption in a low energy region is difficult. On the other hand, in the low energy region where an absorption line due to ²³⁵ U or the like exists in the transmission spectrum of neutron rays, the absorption due to ¹⁰ B appears not in a narrow range like the absorption line but in a wide range. For this reason, ¹⁰ B is difficult to detect by the first conventional technique, but has a great influence on the analysis of ²³⁵ U or the like.

On the other hand, even when ¹⁰ B is detected in the second conventional technique, there is an influence of neutron absorption by ²³⁵ U. Further, in general, molten nuclear fuel often contains ¹³⁷ Cs, which is also affected. ¹³⁷ Cs emits 661 keV gamma rays by neutron capture. This energy is substantially equal to the Compton edge when 661 keV gamma rays are detected, although it is far from the 478 keV gamma rays emitted by ¹⁰ B.

Here, the Compton edge will be described below. FIG. 6 is an example of a detection spectrum obtained when the γ-ray detector detects 661 keV γ-rays emitted from ¹³⁷ Cs. In general, it is not γ-rays but directly detected particles (electron or electron-hole pairs) generated by absorption of γ-rays or visible light generated by the γ-ray detector. When detecting γ-rays of a certain energy with a γ-ray detector, the total energy of the γ-rays is transferred to charged particles or visible light, and one γ-ray photon to be detected It is recognized as a pulse signal generated by light. For this reason, the count number of this pulse becomes the count number of γ rays. On the other hand, the pulse height of this pulse is proportional to the number of charged particles and visible light, which corresponds to the energy of incident γ rays. As a mechanism for transferring the energy of γ rays to charged particles (electrons), for example, Compton scattering is performed, and by performing this multiple times, the total energy of γ rays is converted into electrons in a γ ray detector (eg, scintillator). May be transferred to other energy sources. The sharp peak at 0.661 MeV in FIG. 6 corresponds to this case.

However, only a part of the γ-ray energy is transferred to the electron energy, and the γ-rays after passing the energy escape to the outside of the γ-ray detector. This γ-ray may be recognized as a corresponding pulse signal. An output at an energy lower than 661 keV in FIG. 6 corresponds to this case. In practice, Compton scattering is performed only once, and thus γ-rays that have lost energy often escape to the outside of the γ-ray detector. Moreover, the maximum value of the energy that γ-rays lose by one Compton scattering is determined. For this reason, in the γ-ray energy corresponding to this maximum energy, a discontinuous edge whose detection efficiency changes greatly occurs. This is the Compton edge, and the energy of the Compton edge is determined only by the energy of the original γ-ray and the mass of the electron colliding with it, and is about 470 keV for the 661-keV γ-ray. As shown in FIG. 6, the number of counts immediately below the Compton edge is very large. Therefore, in the presence of ¹³⁷ Cs that emits 661 KeV gamma rays, it is difficult to correctly recognize the 478 keV gamma rays emitted by ¹⁰ B.

Therefore, detection of 478 keV gamma rays emitted by ¹⁰ B is greatly affected by the presence of ¹³⁷ Cs. That is, also in the second prior art, the detection accuracy of ¹⁰ B is strongly influenced by the presence of other nuclides that are present simultaneously.

  That is, it has been difficult to accurately perform composition analysis of nuclides in a non-contact and non-destructive manner in a sample in which many nuclides are mixed.

  The present invention has been made in view of such problems, and an object thereof is to provide an invention that solves the above problems.

In order to solve the above problems, the present invention has the following configurations.
The nuclide composition analyzer of the present invention is a nuclide composition analyzer that performs composition analysis of nuclei contained in a sample by irradiating the sample with a neutron beam, and generates a non-monochromatic pulsed neutron beam and irradiates the sample A plurality of time points starting from the generation time of the pulsed neutron beam, from the output of the γ-ray detection unit, the neutron source that performs, a γ-ray detection unit that detects neutron capture γ-rays generated from the sample By calculating the composition of the nuclide that emitted the neutron capture γ-ray at each time point by calculating the intensity of the neutron capture γ-ray, and by statistically processing the composition of the nuclide calculated at each time point, And a data processing unit for calculating the composition of the nuclide.
The nuclide composition analyzer of the present invention includes a neutron detector that detects a transmitted neutron beam after the neutron beam has passed through the sample, and the data processor starts from the generation time of the pulsed neutron beam From the neutron resonance absorption in the spectrum of the transmitted neutron beam detected by the TOF (Time of Flight) method, the composition of the nuclide different from the nuclide whose composition is calculated using the output of the γ-ray detector is calculated. It is characterized by.
In the nuclide composition analyzer of the present invention, the γ-ray detection unit is installed on the upstream side in the incident direction of the γ-ray, and a shield that selectively allows the γ-ray to pass through an opening provided in a part thereof. A scintillator that emits light by absorbing γ-rays, a light emission detector that detects the light emission downstream of the incident direction of the γ-rays, and a light emission that absorbs γ-rays between the scintillator and the shield. And a back catcher scintillator provided with an opening for allowing the γ-rays that have passed through the opening to pass therethrough.
In the nuclide composition analyzer of the present invention, the scintillator and the back catcher scintillator are made of a material mainly composed of LaBr ₃ .
The nuclide composition analysis method of the present invention is a nuclide composition analysis method for performing composition analysis of nuclei contained in a sample by irradiating the sample with a neutron beam, which generates a pulsed neutron beam that is not monochromatic and irradiates the sample A neutron irradiation step, a plurality of time points starting from the generation time of the pulsed neutron beam, a γ-ray detection step for detecting neutron capture γ-rays generated from the sample, and each of the plurality of time points Temporarily calculating the composition of the nuclide that emitted the neutron capture γ-ray at each time point by calculating the intensity of the neutron capture γ-ray at the time point, and statistically processing the composition of the nuclide temporarily calculated at each time point And a first composition calculation step for calculating the composition of the nuclide.
In the nuclide composition analysis method of the present invention, the nuclide whose composition is calculated in the first composition calculation step is at least one of ¹ H, ¹⁰ B, ⁵⁶ Fe, ⁵³ Cr, ⁵⁸ Ni, and ⁹⁵ Mo. It is characterized by.
In the nuclide composition analysis method of the present invention, a spectrum of a transmitted neutron beam after the neutron beam has passed through the sample is detected by a TOF (Time of Flight) method starting from the generation time of the pulsed neutron beam. A neutron detection step, and a second composition calculation step for calculating a composition of a nuclide different from the nuclide whose composition is calculated in the first composition calculation step from neutron resonance absorption in the spectrum of the transmitted neutron beam. Features.
In the nuclide composition analysis method of the present invention, the nuclide whose composition is calculated in the second composition calculation step is ²³⁵ U, ²³⁸ U, ²³⁸ Pu, ²³⁹ Pu, ¹⁴⁰ Pu, ²⁴¹ Pu, ²⁴² Pu, ¹³¹ Xe, ¹³³ Cs. , ¹⁵² Sm.

  Since the present invention is configured as described above, the composition analysis of nuclides can be performed accurately in a non-contact and non-destructive manner in a sample in which many nuclides are mixed.

It is a figure which shows the structure of the nuclide composition analyzer used as embodiment of this invention. It is a figure which shows the neutron energy dependence of the neutron self-absorption correction factor in a sample. It is sectional drawing which shows the structure of the gamma-ray detector used for the nuclide composition analyzer used as embodiment of this invention. It is a spectrum obtained when ¹³⁷ Cs neutron capture gamma rays are detected by the gamma ray detector of the example and the gamma ray detector of comparative example 1. It is an example of the spectrum of the transmission neutron beam obtained by the neutron detection process. It is a figure explaining generation | occurrence | production of the Compton edge in a gamma ray detector.

  Hereinafter, the configuration of the nuclide composition analyzer and the nuclide composition analysis method according to the embodiment of the present invention will be described. The nuclides detected here are diverse, and even when a sample contains many nuclides, the composition analysis of each nuclide can be performed precisely.

  Here, the sample is irradiated with a neutron beam, γ rays emitted from the sample are detected, and the spectrum of the transmitted neutron beam is also measured. This neutron beam has a non-monochromatic pulsed output. Moreover, the nuclide whose composition analysis is performed from the measurement result of γ-ray is different from the nuclide whose composition analysis is performed from the measurement result of transmitted neutron. Here, Table 1 shows the former nuclide and the nuclear reaction equation used for its detection, and Table 2 shows the latter nuclide and the nuclear reaction equation used for its detection.

  It is the value of neutron resonance energy that greatly differs between the nuclides in Table 1 and the nuclides in Table 2. The neutron resonance energies of the nuclides in Table 2 are all smaller than those in Table 1 and are 40 eV or less. The neutron resonance energy of the nuclides in Table 1 is 45 eV even with the smallest 45 Mo, and all others are 1 keV or more. As will be described later, the TOF method (Time of Flight method) is used to examine the spectrum of a transmitted neutron beam. However, it is difficult to accurately measure the absorption line by the TOF method for such high energy neutrons. It is. For this reason, it becomes difficult to investigate the absorption line in the transmission neutron beam as described in Patent Document 1 for the nuclides described in Table 1. For this reason, it is appropriate to use the detection results of neutron capture γ rays for the composition analysis of the nuclides in Table 1. At this time, since the neutron beam is pulsed and its spectrum is not monochromatic, the composition can be calculated for each different neutron energy.

  On the other hand, composition analysis for the nuclides in Table 2 can be performed by analyzing the spectrum of the transmitted neutron beam. This is the same as the first prior art. At this time, since the composition of the nuclides in Table 1 is obtained with high accuracy, the composition analysis for the nuclides in Table 2 can be performed with high accuracy by using this data.

  FIG. 1 is a diagram showing the configuration of the nuclide composition analyzer 10. In the neutron source 11, the (quasi) monochromatic pulsed fast neutrons emitted from the neutron generation source 111 pass through the moderator 112 and become incident neutron beams (neutron beams) 20 having a spread in energy. . The incident neutron beam 20 passes through the collimator 12 in which a narrow opening is formed in the optical axis direction to become a thin beam with a limited beam diameter, and is irradiated onto the sample 100 (neutron irradiation step). The neutron capture γ-rays 30 emitted by the nuclei in the sample 100 capturing neutrons in the incident neutron beam 20 are γ provided around the sample 100 (in a direction away from the optical axis of the incident neutron beam 20). It is detected by a line detector (γ-ray detector) 13 (γ-ray detection step). On the other hand, the transmitted neutron beam 21 after the incident neutron beam 20 has passed through the sample 100 is detected by a neutron detector (neutron detector) 14 (neutron detection step).

  The data processing unit 15 is a personal computer, for example, and calculates the composition of nuclides contained in the sample 100 from the outputs of the γ-ray detector 13 and the neutron detector 14.

  Here, as the neutron generation source 111, any source capable of generating controlled pulsed neutrons can be used, but deuterons are accelerated by an accelerator to collide with a tritium target, and a 14 MeV monochrome ( A fusion neutron source emitting quasi-monochromatic neutrons is particularly preferably used. The neutron generation timing (pulse rise timing) in the neutron generation source 111 is recognized by the data processing unit 15. The data processing unit 15 can use the outputs of the γ-ray detector 13 and the neutron detector 14 at a plurality of time points starting from this time for calculating the nuclide composition.

  As the material of the moderator 112, for example, water, heavy water, polyethylene, or the like that is generally used for moderation of fast neutrons can be used. As the material of the collimator 12, for example, a material obtained by alternately combining a plate-like Li or B-containing polyethylene plate and a lead plate can be used. The shape and size of the opening provided in the collimator 12 can be appropriately set according to the shape and the like of the sample 100, thereby limiting the beam size and the like of the incident neutron beam 20 irradiated on the sample 100. The collimator 12 is preferably thicker in the optical axis direction in order to limit the beam size of the incident neutron beam 20 and the like.

Thereafter, the nuclei in the sample 100 irradiated with the incident neutron beam 20 are in an excited state by capturing neutrons, and emit neutron capture γ rays 30 when de-excited. The neutron capture γ-ray 30 is detected by a γ-ray detector 13 provided around the sample 100. For example, when a radionuclide such as ¹³⁷ Cs is included in the sample 100, decay γ rays 31 emitted by the natural decay are also emitted and detected by the γ ray detector 13. For this reason, in order to detect a specific neutron capture γ-ray 30 (recognized in the detection spectrum), the γ-ray detector 13 with high energy resolution is required. The configuration of the γ-ray detector 13 will be described later. On the other hand, a neutron detector 14 for detecting the transmitted neutron beam 21 is provided behind the sample 100 in the optical axis direction of the incident neutron beam 20. As the neutron detector 14, for example, a combination of a scintillator and a photomultiplier tube can be used. In this case, it is preferable to use a scintillator having a low γ-ray detection efficiency and a high neutron detection efficiency.

  Here, the role of the γ-ray detector 13 is the same as that in the second prior art, and the role of the neutron detector 14 is the same as that in the first prior art. However, in this nuclide composition analyzer 10, the incident neutron beam 20 controlled in a pulse shape is used, and the γ-ray detector 13 and the neutron detector 14 are simultaneously used. It is possible to perform composition analysis of the nuclide that has been performed with high accuracy. This will be described below.

  First, since the pulsed incident neutron beam 20 is used here, the generation timing thereof is strictly controlled. In addition, the incident neutron beam 20 is formed when the quasi-monochromatic fast neutrons formed by the neutron source 111 pass through the moderator 112, so that the spectrum is not monochromatic but has a broad energy. For this reason, the detection timing of the neutron in the neutron detector 14 after the generation timing in the neutron source 11 depends on the energy (velocity) of the neutron. That is, since the time dependence of the measurement result in the neutron detector 14 is directly dependent on the energy of neutrons, the energy spectrum of the transmitted neutron beam 21 can be measured by a so-called TOF (Time of Flight) method. This is the same as the first prior art.

However, here, the timing at which the neutron capture γ-ray 30 is emitted from the sample 100 is also determined according to the energy of the neutron in the incident neutron beam 20. That is, the time dependency of the detection result of the γ-ray detector 13 corresponds to the neutron energy dependency of the neutron capture γ-ray 30. The neutron energy dependence of the neutron capture γ-ray 30 in this case will be described below. Here shows the results calculated for ^{10 B (n, αγ) 7} γ -rays emitted case where reaction occurred indicated by Li when the incident neutron beam 20 to ^{10 B} in Table 1 is irradiated. Sample 100 has the composition shown in Table 3 assuming a molten fuel. In this case, the count number of 478 keV neutron capture γ rays detected in the three energy ranges of neutrons ((a) 0.3 to 0.5 eV, (b) 3 to 5 eV, (c) 30 to 50 eV) is calculated. did. Here, the incident flux of neutron was 10 ¹² / cm ² , the peak detection efficiency was 0.01, and the measurement time was 1 day.

Further, the reaction does not occur uniformly in the sample 100, and the number of neutrons that react in the sample 100 is smaller than that on the surface. For this reason, the value obtained by multiplying the above-mentioned count number by a neutron self-absorption correction factor is actually the actual count number (number of γ rays to be detected). ^The results of calculating the neutron energy dependence of the neutron self-absorption correction factor for ¹⁰ B are shown in FIG. The neutron self-absorption correction factor approaches 1 as the energy increases, that is, the influence of self-absorption decreases.

  Here, the number of counts and the neutron self-absorption correction factor (average) in the above three energy ranges are as shown in Table 4.

  That is, the number of γ-rays to be detected increases on the low energy side, but the neutron self-absorption correction factor decreases, and vice versa on the high energy side.

Here, a statistical error in this count number (Poisson (1σ) error of the count number statistic: for example, when the count number of the peak region of 478 keV is A and the corresponding background count number is B, √ (A + B )) Was calculated in consideration of the detection resolution (described later) of the actual γ-ray detector 13 for 478 keV γ-rays in the presence of generation of 10 ⁸ / sec of 661 keV γ-rays caused by ¹³⁷ Cs. The calculation error of the neutron self-absorption correction factor was assumed to be 30% of the value of the neutron self-absorption correction factor. Table 5 shows the results of calculating the statistical error of detecting 478 keV γ-rays in the above three energy regions, the systematic error of the neutron self-absorption correction factor, and the total error calculated from these.

  The results in Table 5 are examples, and the results differ depending on the detection resolution of the γ-ray detector 13 and the error in the neutron self-absorption correction coefficient. However, in any case, the total error may be over 10% and cannot be ignored. In such a case, by using the measurement results in different energy ranges obtained by using the same incident neutron beam 20, the data processing unit 15 increases the intensity of the 478 keV neutron capture γ-ray 30 by various statistical methods. It is clear that it can be calculated with accuracy. Here, as in the TOF method, the detection timing in the γ-ray detector 13 corresponds to the neutron energy.

  For this reason, the neutron capture γ-rays 30 generated from the sample 100 are detected by the γ-ray detector 13 at a plurality of time points starting from the generation time of the incident neutron beam 20 (γ-ray detection step). Next, the composition of the nuclide that emitted the neutron capture γ-ray 30 is temporarily calculated by calculating the intensity of the neutron capture γ-ray at each time point. Next, the composition of the nuclide is calculated by statistically processing the composition calculated at each time point (first composition calculation step). When calculating a more accurate composition from the temporary composition at each time point, fitting or the like in consideration of the above error can be performed. Thereby, it is possible to calculate a more accurate composition.

Next, a specific configuration of the γ-ray detector 13 suitable for this measurement will be described. As described above, when the sample 100 contains ¹³⁷ Cs, the Compton edge of the decayed γ-ray 31 (661 keV) emitted from ¹³⁷ Cs unrelated to the neutron-captured γ-ray is neutron-captured γ-ray (478 keV) emitted from ¹⁰ B. Since they almost overlap, it becomes difficult to identify neutron capture γ rays emitted by ¹⁰ B. In order to improve this point, the γ-ray detector 13 in which the Compton edge is less likely to occur is effective. For this purpose, a configuration in which the possibility that the γ-rays scattered back in the γ-ray detector 13 escape to the outside of the γ-ray detector 13 is effective.

  FIG. 3 is a cross-sectional view showing the configuration of the γ-ray detector 13. In the γ-ray detector 13, as in the neutron detector 14, a scintillator 131 and a photomultiplier tube (luminescence detector) 132 are used. Incident γ-ray 70 (neutron capture γ-ray 30) enters the scintillator 131 through an opening 134 provided in the shield 133. The shield 133 is made of, for example, lead or bismuth as a material that does not transmit the incident γ-ray 70 except the opening 134. In FIG. 3, the shield 133 is formed on the entire surface of the scintillator 131 and the photomultiplier tube 132 in order to shield unnecessary γ-rays that become noise. When negligible, the shield 133 (opening 134) may be provided only in the incident direction of the incident γ-ray 70.

As a material of the scintillator 131, LaBr ₃ (Ce) having a high energy resolution in the energy region of the neutron capture γ-ray 30 is particularly preferably used. LaBr ₃ (Ce) is a material containing LaBr ₃ as a main component and Ce added as a light emitting impurity. Electrons generated by Compton scattering or the like in the scintillator 131 emit light that becomes visible light in the process of decaying its kinetic energy. Therefore, a photomultiplier tube (light emission detector) 132 is provided to detect this light emission. It has been. For this reason, one pulse output is obtained from the photomultiplier tube 132 by one incidence of γ-ray photons, and the count number of these pulses becomes the number of incident photons of γ-rays. In addition, this pulse height is proportional to the number of generated photons of visible light, which is proportional to the number of photoelectrons generated in the photomultiplier tube. Correspond. However, when γ rays that escape to the outside of the scintillator 131 after only a part of the energy is converted to electrons are detected, this correspondence is not established. On the other hand, the upper limit value of the energy that γ rays lose by one Compton scattering is determined. For this reason, at the energy corresponding to this upper limit value, the wave height spectrum of the γ-ray detector 13 undergoes a discontinuous change that greatly decreases. This is the Compton edge in the wave height spectrum shown in FIG. Note that the γ-rays lose a large amount of energy due to Compton scattering (giving the largest energy to the electrons) when the γ-rays are scattered (backscattered) in the incident direction. For this reason, the energy of the Compton edge corresponds to the energy of the γ-ray after scattering when the γ-ray is scattered by changing the direction by 180 °.

  In the γ-ray detector 13, a back catcher scintillator 135 is provided on the incident direction side (between the shield 133 and the scintillator 131) in the scintillator 131 in order to make it difficult to generate a Compton edge. The back catcher scintillator 135 is preferably made of the same material as that of the scintillator 131, and a back catcher opening (opening) 136 is provided at a location corresponding to the opening 134 of the shield 133.

According to this configuration, the incident γ-ray 70 that has passed through the opening 134 of the shield 133 passes through the back catcher opening 136 as it is, reaches the scintillator 131, and emits light there. In the configuration of FIG. 3, γ rays scattered backward by 180 ° cannot be captured, but backscattered γ rays 71 scattered backward by an angle close to 180 ° can be captured again by the back catcher scintillator 135. Here, the light is converted into visible light in the same manner as the scintillator 131, and this light emission can be detected by the photomultiplier tube 132. For this reason, most of the backscattered γ-rays 71 can be captured again by the back catcher scintillator 135, and the influence of the Compton edge can be reduced. FIG. 4 shows the detection spectrum of ¹³⁷ Cs when the γ-ray detector 13 (Example) having this configuration is used and when the shield 133 and the back catcher scintillator 135 are not provided in the same configuration (Comparative Example 1). This is the calculated result. Here, the scintillator 131 is made of LaBr ₃ (Ce) having a diameter of 10 cm and a length of 15 cm, the back catcher scintillator 135 is made of the same material having a diameter of 10 cm and a length of 10 cm, and the opening 134 and the back catcher scintillator opening 136 have a diameter of 2 cm, The thickness of the shield 133 around the opening 134 was 10 cm. With the configuration of the example, it can be confirmed that the Compton edge is reduced to a negligible level. That is, in this case, it is possible to appropriately detect the neutron capture γ-rays 30 emitted by ¹⁰ B even in the presence of ¹³⁷ Cs.

Table 6 shows the ratio of the area of the peak component when 661 keV, which is a decayed γ ray of ¹³⁷ Cs, to the area of the continuous component near 478 keV, which is near the Compton edge, in consideration of the energy resolution of the gamma ray detector. These are the results calculated for the example and comparative example 1 when the energy width when calculating the area is 10 keV and 30 keV, respectively. In addition, as Comparative Example 2, a result obtained when a coaxial Ge detector is used is also shown. Here, the width of 10 keV is the required energy width in the case of the Ge detector, and the width of 30 keV is the required energy width in the case of the LaBr ₃ (Ce) detector.

  From this result, it can be confirmed that high resolution can be obtained in the embodiment using the back catcher scintillator 135. In calculating the values in Table 5, the results when the spectrum width is 30 keV (resolution 548) are used.

As described above, in the nuclide composition analyzer 10 having the configuration of FIG. 1, the composition analysis of ¹⁰ B can be performed with higher accuracy than the output of the γ-ray detector 13. As long as the composition is analyzed from the intensity of the neutron-captured γ-ray, it is clear that the composition analysis can be similarly performed for other nuclides in Table 1. That is, composition analysis can be performed with high accuracy for the nuclides shown in Table 1 included in the sample 100.

  Thereafter, by examining the spectrum of the transmitted neutron beam 21 detected by the neutron detector 14 by the TOF method, the composition analysis for the nuclides shown in Table 2 can be performed. FIG. 5 shows an example in which the spectrum of a transmitted neutron beam 21 having a composition shown in Table 3 and transmitted through a sample having a thickness of 1 cm, 5 cm, and 10 cm is calculated. Here, the spectrum of the incident neutron beam 20 is also shown. The data processing unit 15 can perform the composition analysis of the nuclides shown in Table 2 from the absorption line in this spectrum (second composition calculation step). This is the same as the first prior art. However, in this case, since the composition analysis of the nuclide shown in Table 1 is performed independently of this composition analysis, the result can be used. For this reason, the composition analysis of the nuclide shown in Table 2 can also be performed with high accuracy.

  Further, the data processing unit 15 performs fitting on the measurement result of the γ-ray detector 13 and the measurement result of the neutron detector 14 by using the nuclide composition shown in Table 1 and the nuclide composition shown in Table 2 as parameters. By performing in consideration of the above error, the composition can be calculated with higher accuracy. In this case, the first composition calculation step and the second composition calculation step are performed independently, and the composition value obtained thereby can be used as the initial value of the fitting.

  Thus, according to the nuclide composition analyzer and the nuclide composition analysis method described above, the nuclide composition analysis can be accurately performed in a non-contact and non-destructive manner in a sample in which many nuclides are mixed. As a sample containing a plurality of nuclides described in Tables 1 and 2 at the same time, for example, there is a molten nuclear fuel. When analyzing molten nuclear fuel in various structures, it is particularly preferable to perform this in a non-contact and non-destructive manner. Therefore, the above nuclide composition analyzer and nuclide composition analysis method are suitable.

  In the above example, the neutron detection step and the second composition calculation step were performed using a neutron detector. For example, when it is known in advance that the sample contains only the nuclides shown in Table 1. These are unnecessary. Even in this case, the composition analysis of the nuclides shown in Table 1 can be performed with higher accuracy than the second prior art by the γ-ray detector and the first composition calculation process using the γ-ray detector.

  In the above example, the case where the γ-ray detector having the configuration shown in FIG. 3 is used is described. However, even when the conventional γ-ray detector is used, a plurality of neutron energies are supported. Thus, the provisional composition can be calculated, and the more accurate composition can be calculated by the statistical processing. Therefore, it is clear that the above-described nuclide composition analyzer and nuclide composition analysis method are effective regardless of the configuration of the γ-ray detector.

  Further, as long as the neutron source can generate a pulsed neutron beam that is not monochromatic, it is obvious that the neutron source can be used in the same manner as described above even when a non-fusion neutron source is used. This is because by using a moderator as in the configuration of FIG. 1, the kinetic energy of neutrons can be attenuated to the resonance region and used. However, it is particularly preferable to use a neutron source that can easily control the neutron intensity, the time width of the neutron pulse, and the number of repetitions, such as the above-mentioned fusion neutron source and neutron source using bremsstrahlung by an electron accelerator. .

10 nuclide composition analyzer 11 neutron source 12 collimator 13 γ-ray detector (γ-ray detector)
14 Neutron detector (neutron detector)
15 Data processor 20 Incident neutron beam (neutron beam)
21 Transmitted neutron beam 30 Neutron capture γ-ray 31 Decayed γ-ray 70 Incident γ-ray 71 Back-scattered γ-ray 100 Sample 111 Neutron source 112 Moderator 131 Scintillator 132 Photomultiplier tube (luminescence detector)
133 Shield 134 Opening 135 Back catcher scintillator 136 Back catcher opening (opening)

Claims (8)

A nuclide composition analyzer that performs composition analysis of nuclei contained in a sample by irradiating the sample with a neutron beam,
A neutron source that generates a pulsed neutron beam that is not monochromatic and irradiates the sample; and
A γ-ray detector for detecting neutron capture γ-rays generated from the sample;
The composition of the nuclide that emitted the neutron capture γ-ray is calculated by calculating the intensity of the neutron capture γ-ray from the output of the γ-ray detector at a plurality of time points starting from the generation time of the pulsed neutron beam. Temporarily calculating for each time point, and by statistically processing the composition of the nuclide temporarily calculated at each time point, a data processing unit that calculates the composition of the nuclide,
The nuclide composition analyzer characterized by comprising. Comprising a neutron detector for detecting a transmitted neutron beam after the neutron beam has passed through the sample;
The data processing unit outputs the output of the γ-ray detection unit based on neutron resonance absorption in the spectrum of the transmitted neutron beam detected by the TOF (Time of Flight) method starting from the generation time of the pulsed neutron beam. The nuclide composition analyzer according to claim 1, wherein a composition of a nuclide different from the nuclide whose composition is calculated by using the nuclide is calculated. The γ-ray detector is
a shield that is installed on the upstream side of the incident direction of γ-rays and selectively allows the γ-rays to pass through an opening provided in a part thereof;
a scintillator that emits light by absorbing γ rays;
A light emission detector for detecting the light emission on the downstream side in the incident direction of the γ-ray;
Between the scintillator and the shield, a back catcher scintillator provided with an opening that emits light by absorbing γ-rays and passes γ-rays that have passed through the opening,
The nuclide composition analyzer according to claim 1, wherein the nuclide composition analyzer is provided. The scintillator and the back catcher scintillator nuclide composition analysis apparatus according to claim 3, characterized in that it is composed of a material mainly composed of LaBr _3. A nuclide composition analysis method for performing composition analysis of nuclei contained in a sample by irradiating the sample with a neutron beam,
A neutron irradiation step of generating a pulsed neutron beam that is not monochromatic and irradiating the sample; and
Γ-ray detection step of detecting neutron capture γ-rays generated from the sample at a plurality of time points starting from the generation time of the pulsed neutron beam;
Temporarily calculating the composition of the nuclide that emitted the neutron capture γ-ray at each time point by calculating the intensity of the neutron capture γ-ray at each time point among the plurality of time points, the tentatively calculated at each time point A first composition calculating step of calculating the composition of the nuclide by statistically processing the composition of the nuclide;
The nuclide composition analysis method characterized by comprising. The nuclide whose composition is calculated in the first composition calculating step is at least one of ¹ H, ¹⁰ B, ⁵⁶ Fe, ⁵³ Cr, ⁵⁸ Ni, and ⁹⁵ Mo. Nuclide composition analysis method. A neutron detection step of detecting a spectrum of a transmitted neutron beam after the neutron beam has passed through the sample by a TOF (Time of Flight) method starting from the generation time of the pulsed neutron beam;
A second composition calculating step of calculating a composition of a nuclide different from the nuclide whose composition is calculated in the first composition calculating step from neutron resonance absorption in the spectrum of the transmitted neutron beam;
The nuclide composition analysis method according to claim 5 or 6, characterized by comprising: The nuclide whose composition is calculated in the second composition calculation step is at least one of ²³⁵ U, ²³⁸ U, ²³⁸ Pu, ²³⁹ Pu, ¹⁴⁰ Pu, ²⁴¹ Pu, ²⁴² Pu, ¹³¹ Xe, ¹³³ Cs, and ¹⁵² Sm. The nuclide composition analysis method according to claim 7, wherein:

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