JP2024030112A - Uranium radiation measurement method - Google Patents

Abstract

To provide a method that measures radiation of refined uranium under coexistance of refined uranium and NORM uranium in waste discharged from a facility handling a nuclear fuel material or nuclear raw material.SOLUTION: A measurement method of the present invention includes the steps of: (A) preparing a measurement vessel provided with an opening part, a bottom part and a lateral part with a prescribed depth; (B) arranging a NaI(Tl) scintillation radiation detector on an outer side of the lateral part of the measurement vessel; (C) measuring a γ-ray spectrum from waste accommodated in the measurement vessel by the NaI(Tl) scintillation radiation detector; (D) subtracting peaks of adjacent and interfering 0.934MeV and 1.120MeV of 214Bi from a peak in 1.001MeV after subtracting a background from the γ-ray spectrum, and discriminating a peak of 1.001MeV 234mPa generating in a collapse process of 238U discharges; and (E) calculating radiation of refined uranium from the discriminated peak of the 1.001MeV.SELECTED DRAWING: Figure 1

Description

The present invention generally relates to a method for measuring the radioactivity of uranium, and specifically relates to a method for measuring the radioactivity of uranium in waste generated from nuclear facilities, etc. The present invention relates to a method for measuring the radioactivity of purified uranium in order to distinguish between purified uranium that may be contained therein and naturally derived uranium contained in general civil engineering and industrial materials. Technical field

In addition to radioactive waste that requires special management from the perspective of radiation protection in facilities that handle nuclear fuel materials or nuclear raw materials such as nuclear facilities, waste that is not radioactive waste (hereinafter referred to as "NR waste") ) also occur in large quantities. Regarding this NR waste, the Ministry of Economy, Trade and Industry has issued guidelines (instructions) regarding its handling (Non-Patent Document 1), and it is necessary to judge it as waste and manage it in accordance with the contents. For example, materials installed in controlled areas and articles used in controlled areas that may be determined to be NR waste are required to undergo radiation measurement evaluation just to be sure.

In addition, if concrete scraps, etc. generated during the demolition of a building such as a nuclear facility are found to be uncontaminated by primary decontamination measurements, and if no problems were originally identified in the confirmation inspection by the Ministry of Education, Culture, Sports, Science and Technology after primary decontamination, then the building The controlled area was lifted and it fell outside the scope of regulations under the Nuclear Reactor Regulation Act (Reactor Regulation Act), and could be disposed of as building demolition waste (industrial waste). However, if building demolition waste such as concrete scraps is managed within the regulatory framework of the Reactor Regulations Act, a radiation measurement evaluation is required to ensure safety and security. .

Furthermore, items (including waste) taken out of controlled areas must be checked to be free of contamination in accordance with other regulations, such as the Ionizing Radiation Hazard Prevention Ordinance (Ministry of Labor Ordinance No. 41 of September 30, 1971). It is necessary to carry out transportation etc. after checking.

Judgment as to whether or not waste falls under NR waste at facilities that handle uranium, such as nuclear facilities (hereinafter also referred to as "NR judgment"), is based on waste that has been contaminated, with the contaminated parts clearly separated. The judgment is based on the fact that there is no clear history. The measurements taken when determining NR are to measure all the uranium in the waste, just to be sure that there is no contamination. Confirmation that there is no contamination can be based on, for example, that uranium is below the theoretical detection limit.

However, this total uranium includes not only purified uranium used in research, but also naturally occurring uranium (NORM), which is contained in civil engineering materials such as concrete, soil, and abrasive materials. ) may be mixed. At that time, if NORM uranium is detected even if there is no remaining refined uranium, it will be necessary to either recover the uranium in the United States or dispose of it in domestic burial, increasing the amount and cost of processing. There is a risk that it will happen.

Furthermore, when measuring to be sure about purified uranium contamination, it is not possible to determine whether there is a possibility that purified uranium remains based only on the measurement results for total uranium. However, if purified uranium and NORM uranium can be distinguished and it is confirmed that purified uranium is below the detection limit (approximately 0.05 to 0.1 Bq/g), it can be shown that there is no residual contamination by purified uranium. This is effective in terms of safety management at the facility. In addition, measurements were taken to make sure that there was no generated furan other than NORM, and by placing it outside the scope of the Reactor Regulation Act, it was possible to dispose of it as general waste in the same way as building dismantled waste after the control zone was lifted. Become.

The above-mentioned building demolition waste such as concrete debris is usually stored in storage containers such as drums, and chemical analysis can be performed after sampling from the storage container. However, in that case, ICP-MASS analysis of the isotopes of normal uranium ( ²³⁸ U and ²³⁵ U) will be performed as a direct measurement, and if NORM uranium and purified uranium are natural uranium ( ²³⁸ U and ²³⁵ U), then both will be the same. Cannot be identified. It is also possible to use a method in which samples are collected from the storage container and analyzed using a NaI (Tl) scintillation radiation detector (hereinafter also simply referred to as a NaI radiation detector) or a Ge semiconductor radiation detector. . However, in this case, since the contents of the storage container are not homogenized, it is not possible to obtain measurement results for the entire contents of the storage container.

As NORM uranium, there are those that are in a permanent equilibrium state with uranium as the parent nuclide, those that are purified uranium as the starting nuclide, or those that are in a permanent equilibrium state with ²²⁶ Ra as the starting nuclide due to some manipulation etc. is assumed. When each gamma ray spectrum is measured using a NaI radiation detector, almost the same spectra are obtained. The difference is the presence or absence of a slight γ-ray (1.001 MeV) emitted from ²³⁴ Th and ^{234 mPa} after ²³⁸ U decay. Purified uranium emits gamma rays (1.001 MeV) from ^{234 mPa} . Therefore, in order to evaluate the radioactivity of purified ²³⁸ U, it is necessary to discriminate the peak of the ^{234 mPa} gamma ray (1.001 MeV) spectrum.

Patent Document 1 discloses a method for non-destructively measuring the ^amount of uranium ( ²³⁵ U and ²³⁸ U) contained in radioactive waste stored in a container from the outside of the container ^. A method for determining the 1.001 MeV gamma ray intensity emitted by Pa from the gamma ray spectrum measured with a NaI radiation detector is disclosed.

However, the method of Patent Document 1 detects the amount of uranium in building demolition waste that is not subject to radioactive waste such as concrete scraps, particularly the trace amount of purified uranium in the coexistence of purified uranium and NORM uranium. Not in that way. In addition, in the method of Patent Document 1, radioactive waste for calibration with a known uranium content is first prepared based on the uranium analysis value obtained by destructive measurement of radioactive waste, and the radioactive waste for calibration is injected with the present invention. A calibration curve is created by applying the non-destructive measurement method, and the amount of uranium is calculated using the calibration curve.

JP 5-340861

Regarding the handling of "non-radioactive waste" at nuclear facilities (instructions), Ministry of Economy, Trade and Industry, Nuclear and Industrial Safety Agency, NISA-111a-08-1, May 27, 2008

An object of the present invention is to provide a method for measuring the radioactivity of purified uranium in the coexistence of purified uranium and NORM uranium in waste generated from facilities that handle nuclear fuel materials or nuclear source materials.

One aspect of the present invention provides a measurement method for determining the radioactivity of purified uranium in the coexistence of purified uranium and naturally occurring uranium in waste generated from a facility that handles nuclear fuel material or nuclear source material. The measurement method is
(A) preparing a measurement container having an opening, a bottom, and sides of a predetermined depth;
(B) arranging a NaI (Tl) scintillation radiation detector outside the side of the measurement container;
(C) measuring the γ-ray spectrum from the waste stored in the measurement container with a NaI (Tl) scintillation radiation detector;
(D) After subtracting the background from the γ-ray spectrum, the peak at 1.001 MeV is subtracted from the adjacent and interfering peaks at ^0.934 MeV and 1.120 MeV of 214 Bi, which is generated in the decay process of ²³⁸ U. Discriminating the 1.001 MeV peak emitted by ^{234 mPa} ;
(E) Calculating the radioactivity of purified uranium from the discriminated 1.001 MeV peak.

In one aspect of the present invention, the step (D) of discriminating the 1.001 MeV peak includes:
(D1) obtaining a corrected γ-ray spectrum by subtracting a thorium series standard spectrum corrected according to the γ-ray measurement in the measurement container from the γ-ray spectrum;
(D2) Subtract the peaks at 0.934 MeV and 1.120 MeV from the peak at 1.001 MeV in the corrected γ-ray spectrum to discriminate the 1.001 MeV peak emitted by ^{234 mPa} generated in the decay process of ²³⁸ U. and a step of doing so.

In one aspect of the present invention, the step (D1) of obtaining a corrected γ-ray spectrum includes:
(D10) measuring γ-rays from a predetermined thorium-series γ-ray source and subtracting the background to obtain a thorium-series spectrum;
(D20) Performing density correction in the measurement container on the obtained thorium series spectrum to obtain a thorium series standard spectrum.

In one aspect of the present invention, the step (D) of discriminating the 1.001 MeV peak includes:
(D3) calculating the standard deviation (σ) when fitting the total absorption peak of each energy of the spectrum data obtained by gamma ray measurement of the gamma ray standard source to a normal probability distribution;
(D4) From the net counts obtained from 1.765 MeV (15.4%) of ²¹⁴ Bi, the actual interfering peaks of 0.934 MeV (3.03%) and 1.120 MeV (15.1%) are determined. a step of calculating a simulated normal probability distribution (standard deviation (σ) at each peak);
(D5) From the peak at 1.001 MeV in the corrected γ-ray spectrum, calculate the normal probability distribution that simulates the actual peaks at 0.934 MeV (3.03%) and 1.120 MeV (15.1%). subtracting to obtain a peak of 1.001 MeV (0.84%).

In one aspect of the present invention, the step (C) of measuring a γ-ray spectrum from waste includes:
(C1) Different KCl aqueous solutions were placed in a measurement container, and the standard efficiency ⁽ η _E1.461 , unit: cps/(photon/cm ³ )),
(C2) determining the energy relative efficiency (R _η ) of the NaI (Tl) scintillation radiation detector for γ rays;
(C3) using a cylindrical volumetric radiation source model to obtain a density correction coefficient (D _W · D _C ) for the standard efficiency when the density changes for water and concrete;
(C4) Measurement efficiency ηE _x (cps/(photon/cm ³ )) in the measurement container, standard efficiency (η _E1.461 ), energy relative efficiency (R _η ), and density correction coefficient D _W · D _C Using the following formula η _Ex = η _E1.461 ×Rη×D _W・D _C
and a step of determining from.

1 is a diagram showing a basic flow of a measurement method according to an embodiment of the present invention. FIG. 2 is a diagram showing the arrangement of a measurement container and a radiation measuring device according to an embodiment of the present invention. It is a figure showing the calculation flow of measurement efficiency of one embodiment of the present invention. It is a figure showing the density correction coefficient of one embodiment of the present invention. It is a figure showing measurement efficiency of one embodiment of the present invention. FIG. 3 is a diagram showing a processing flow of a γ-ray spectrum according to an embodiment of the present invention. It is a figure showing a thorium series standard spectrum of one embodiment of the present invention. It is a figure showing a uranium series standard spectrum of one embodiment of the present invention. FIG. 2 is a diagram showing a processing flow for discriminating a 1.001 MeV peak emitted by ^{234 m} Pa according to an embodiment of the present invention. FIG. 3 is a diagram showing the standard deviation (σ) when the total absorption peak of each energy of the γ-ray spectrum data of one embodiment of the present invention is fitted to a normal probability distribution. It is a figure showing the normal probability distribution of 0.934 MeV (3.03%) and 1.120 MeV (15.1%) calculated by one embodiment of the present invention. FIG. 3 is a diagram showing the discrimination results of a peak of 1.001 MeV emitted by ^{234 mPa} in gamma ray measurement of a gypsum board that does not contain ²³⁸ U according to an embodiment of the present invention. FIG. 2 is a diagram showing the discrimination results of a peak of 1.001 MeV emitted by ^{234 mPa} in gamma ray measurement of a natural uranium ore containing ²³⁸ U according to an embodiment of the present invention.

Embodiments of the present invention will be described with reference to the drawings. FIG. 1 is a diagram showing the basic flow of an inspection method according to an embodiment of the present invention. In step S1, a measurement container including an opening, a bottom, and a side having a predetermined depth is prepared. In step S2, a NaI (Tl) scintillation radiation detector is placed outside the side of the measurement container. Step S2 may be performed simultaneously and integrally with the preparation of the measurement container in step S1. In step S3, the γ-ray spectrum from the measurement object (NR waste, etc.) in the measurement container is measured with a radiation detector. In step S4, the peaks at 0.934 MeV and 1.120 MeV of adjacent and interfering ²¹⁴ Bi are subtracted from the peak at 1.001 MeV after subtracting the background from the γ-ray spectrum, and the peaks generated in the decay process of ²³⁸ U are calculated. Discriminate the 1.001 MeV peak emitted by ^{234 mPa} . In step S4, the radioactivity of purified uranium is calculated from the discriminated 1.001 MeV peak.

FIG. 2 is a diagram showing the arrangement of a measurement container and a radiation measuring device according to an embodiment of the present invention. FIG. 2 shows a cylindrical measurement container 1 and a radiation detector 3 inserted into a radiation shield 2 located outside the side wall of the measurement container 1. The measurement container 1 includes a cylindrical storage portion with an approximately circular or elliptical opening. The measurement container 1 is made of, for example, a metal or resin drum. The measurement container 1 contains an aqueous solution of potassium chloride (KCl) to be measured, powdered waste (NR waste), shredded waste, compressible waste, flexible waste, etc. It can be filled (stored).

The radiation shield 2 contains a heavy metal such as lead and blocks background radiation from entering the radiation detector 3 from the outside. The radiation shield 2 has one side wall connected to a movable part of the heavy machine 5. The radiation shield 2 has a cylindrical shape with a circular or semicircular opening, and the radiation detector 3 can be taken in and out through the opening. By operating the handle 6 of the heavy equipment 5, the radiation shield 2 can be moved up and down along the side wall of the measurement container 1. This allows the measurement position of the radiation detector 3 within the radiation shield 2 to be changed.

The number of radiation detectors 3 is not limited to one, and two or more can be provided. In this case, for example, a plurality of radiation shields 2 containing one radiation detector 3 may be arranged, or two or more radiation detectors 3 may be arranged within the elongated radiation shield 2. The radiation detector 3 has a built-in detection section 4 at its tip. As the radiation detector 3, for example, an Na(Tl) scintillation detector with a detection portion having a crystal size of Φ63×160 mm can be used. The crystal size is about 3 times that of 63x63, and it is expected that the detection accuracy will improve to the level of accuracy that corresponds to NR measurement.

The radiation detector 3 is connected to a spectrum analyzer (not shown) and a calculation device (not shown) via a communication cable (not shown). The communication cable includes, for example, a GP-IB cable, a USB cable, or a LAN cable. The spectrum analyzer measures the spectrum of γ-rays detected by the radiation detector 3 from the measurement object in the measurement container 1 . The calculation device calculates various parameters (measurement efficiency, radioactivity (density), etc.) for obtaining the radioactivity of uranium in the object to be measured from the measured γ-ray spectrum. The arithmetic device includes, for example, a personal computer (PC).

The waste filled in the measurement container 1 refers to waste that is subject to NR waste generated in facilities that handle nuclear fuel materials and the like. The waste includes powdered waste, shredded waste, compressible waste, flexible waste, etc. that can be filled into the measurement container 1. Powder-like waste in powder form includes not only powder as fine as grains of sand or rice grains, but also particles on the order of several millimeters (<about 10 mm) in size. The waste may include materials such as, for example, flakes of paint or rusted steel, metal, pebbles, concrete pieces, wood, paper, plastic, polysheet, vinyl, rubber, and the like.

Nuclear fuel materials include uranium, thorium, and their compounds, which are sources of α, β, and γ rays. Facilities that handle nuclear fuel materials include nuclear facilities as well as facilities that are not directly related to nuclear power, such as facilities that handle refined uranium, chemical factories that use uranium as a catalyst, and other manufacturing and processing facilities. It will be done. Nuclear facilities include, for example, refining facilities, nuclear reactor facilities, reprocessing facilities, etc. described in Non-Patent Document 1. Furthermore, NR waste basically refers to waste that is judged in accordance with the guidelines in Non-Patent Document 1, but facilities that generate target waste include nuclear facilities as well as facilities that are not directly related to nuclear power as mentioned above. It also includes chemical factories, etc., which are not related to the above.

Next, details of the gamma ray spectrum measurement and its processing flow in steps S3 to S5 in FIG. 1 will be described below with reference to FIGS. 3 to 14 and Tables 1 to 4.

FIG. 3 is a diagram illustrating a flowchart for calculating measurement efficiency when a drum is used as the measurement container 1, taking into account self-absorption due to differences in the weight and material of the stored items to be measured. In step S10, different KCl aqueous solutions are put into the measurement container 1 (drum), and the total absorption peak area count rate (η _E1.461 , ^unit : cps/(photon /cm ³ )) and set it as the standard efficiency (density: 1, material: water, E _γ : 1.461 MeV).

Table 1 shows information regarding the concentration of the KCl aqueous solution used in the test. In the test, KCl aqueous solutions with four different radioactivity concentrations (Bq/cm ³ ) from 0 to 1.419 were prepared and measured.

Table 2 shows the determined standard efficiency, that is, the total absorption peak area count rate (η _E1.461 , unit: cps/(photon/cm ³ )) of γ-rays at 1.461 MeV.

In step S11, the energy relative efficiency (R _η ) based on the 1.461 MeV γ-ray of the NaI (Tl) scintillation radiation detector used is determined. Table 3 shows the determined energy relative efficiency (R _η ).

In step S12, a density correction coefficient (D _W ·D _C ) for the standard efficiency when the density changes is calculated for water and concrete using a cylindrical volume radiation source model. Calculations are performed based on, for example, the cylindrical source calculation method found in the 1968 Engineering Compendium on Radiation Shielding IAEA. FIG. 4 shows the calculation results of the material and the density correction coefficient, with density 1, γ-ray energy 1.461 MeV, and water as the storage material. The calculation results are expressed by a cubic function approximation formula with weight as a variable for each energy. The total weight of the drum is 178 kg.

In step S13, the measurement efficiency ηE _x (cps/(photon/cm ³ )) in the measurement container (drum) is calculated using the reference efficiency (η _E1.461 ), the energy relative efficiency (R _η ), and the density correction coefficient D _W・Using D _C , the following formula

η _Ex = η _E1.461 ×Rη×D _W・D _C

Find from. FIG. 5 shows the calculation results of the measurement efficiency η _Ex (cps/(photon/cm ³ )).

Next, a method for discriminating total absorption peaks at evaluation target energies in overlapping gamma ray spectra in a NaI (Tl) scintillation radiation detector will be described. When uranium and thorium series radioactive substances are mixed, the gamma ray spectrum obtained by measurement will have many broad peaks overlapping each other. When performing analysis, it is necessary to remove these influences, separate the gamma rays of the energy to be evaluated into individual peaks, and calculate the area. Therefore, it is necessary to perform evaluation according to the processing flow of the γ-ray spectrum according to the embodiment of the present invention shown in FIG. In addition, if several energy peaks interfere even with a single series of radioactive substances, data processing is performed using the γ-ray spectrum processing flow of an embodiment of the present invention shown in FIG. 9, which will be described later. It is necessary to treat it as an energy peak.

In step S21 of the γ-ray spectrum processing flow according to the embodiment of the present invention in FIG. 6, γ-ray spectra in a state of radiation equilibrium are measured using uranium and thorium series radiation sources whose self-absorption can be ignored, Subtract the background to obtain standard spectra for uranium and thorium. FIG. 7 shows the obtained thorium series standard spectrum. Figure 8 shows the obtained uranium series standard spectrum. In step S21, the standard spectrum of thorium is adjusted to the 2.615 MeV peak of ²⁰⁸ Tl in the gamma ray spectrum of the waste, and at the same time, a thorium series standard spectrum is calculated by taking into account the density correction coefficient.

In step S22, the thorium series standard spectrum calculated in step S21 is subtracted from the gamma ray spectrum of the waste in the measurement container obtained in step S3 of FIG. 1 to remove the thorium series peak. In step S23, the gamma ray spectra of the waste from which the thorium series or uranium series peaks have been removed in step S22 are each used as corrected gamma ray spectra.

Next, a method for evaluating the concentration of purified uranium when purified uranium is mixed into the uranium series will be explained. Among the radioactive substances treated as NORM, the uranium series includes those in a permanent equilibrium state with uranium as the parent, those in which purified uranium is the starting nuclide, and those in permanent equilibrium with ²²⁶ Ra as the starting nuclide. It is assumed that "things that are in a state" are assumed. In order to evaluate the radioactivity of purified ²³⁸ U, it is necessary to extract 0.934 MeV (3.03%) and 1.0 MeV of adjacent and interfering ²¹⁴ Bi from the corrected gamma-ray spectrum obtained by removing the thorium series peaks. It is necessary to subtract the 120 MeV (15.1%) peak to discriminate the 1.001 MeV (0.84%) peak emitted by ^{234 mPa} generated in the decay process of ²³⁸ U. Note that similar processing can be performed on peaks at other energies. FIG. 9 shows such a processing flow.

In step S30 of FIG. 9, the standard deviation (σ) is determined by fitting the total absorption peak of each energy of the line spectrum obtained by the standard radiation source to a normal probability distribution. The total absorption peak of the γ-ray spectrum was obtained using standard point sources of ¹³⁷ Cs (661 keV) and ⁶⁰ Co (1173 keV, 1333 keV), and the background was cut out using trapezoidal approximation and fitted with a normal probability distribution using the least squares method. The standard deviation (σ, representing the spread of the peak) obtained when doing this is shown in FIG.

In step S31, the relative efficiency of the target gamma rays (1.765 MeV, 0.934 MeV, 1.120 MeV) is determined from the weight and material of the waste in the measurement container. In step S32, the actual interfering peaks of 0.934 MeV (3.03%) and 1.120 MeV (15.1%) are determined from the net counts determined from 1.765 MeV (15.4%) of ²¹⁴ Bi. Calculate the normal probability distribution (standard deviation (σ) at each peak) that simulates

Specifically, for example, the normal probability distribution is calculated as follows.
(1) The peak area (count number) C of ²¹⁴ Bi at 1.765 MeV is corrected by the γ-ray emission rate ratio of ²¹⁴ Bi. That is, for the peak area (count number) C of 1.765 MeV, (0.934 MeV emission rate/1.765 MeV emission rate) = 3.03%/15.4% and (1.120 MeV emission rate/1.765 MeV emission rate) Multiply by 15.1%/15.4% to obtain C ₁ and C ₂ .
(2) Correction is performed based on the gamma ray detection efficiency ratio of the NaI radiation detector. The gamma ray detection efficiency ratio is calculated in advance from the measurement results of the radiation source for calibration. Specifically, the γ-ray detection efficiency ratio (1.48 _(0.934MeV) /0.86 _(1.765MeV) , 1.39 _(1.120MeV) /0.86 _{(1 .765MeV)} ) is multiplied by C ₁ and C ₂ obtained in (1) to obtain C ₁₁ and C ₂₂ .
(3) Correction is performed using the γ-ray self-absorption rate ratio depending on the density of waste in the measurement container, which is the measurement target. The γ-ray self-absorption rate ratio is calculated in advance using a cylindrical source model calculation. Specifically, for example, in the case of a density of 1.27 g/cm ³ , the γ-ray self-absorption rate ratio (0.682 _{(0.934 MeV)} /0.897 _{(1.765 MeV)} and 0.744 ₍₁ C 11 and _{C 22 obtained} in (2) are multiplied by C ₁₁ and C ₂₂ obtained in (2) to obtain C ₁₁₁ and C ₂₂₂ . C ₁₁₁ and C ₂₂₂ correspond to the simulated peak areas (count numbers) of 0.935 MeV and 1.120 MeV.

FIG. 11 shows a normal probability distribution simulating the calculated actual peaks of 0.934 MeV (3.03%) and 1.120 MeV (15.1%). In FIG. 10, the lower spectrum A is the calculated normal probability distributions C ₁₁₁ and C ₂₂₂ , and the upper spectrum B is the spectrum of the uranium series actually measured with a Ge semiconductor detector. It can be seen that the calculated spectrum A (C ₁₁₁ , C ₂₂₂ ) closely approximates the actual spectrum B.

In step S33, from the peak at 1.001 MeV in the corrected γ-ray spectrum obtained in step S24 of FIG. %) to obtain a peak of 1.001 MeV (0.84%).

Based on the measured spectra of gypsum board, which was found to be in radioactive equilibrium with nuclides of ²²⁶ Ra and above and does not contain ²³⁸ U, using a Ge semiconductor detector in advance, and natural uranium ore, which contains ²³⁸ U. , an experiment was conducted to obtain a peak of 1.001 MeV by subtracting the normal probability distribution that simulated the actual peaks of 0.934 MeV and 1.120 MeV calculated in step S32.

FIG. 12 is a diagram showing the discrimination results of a peak of 1.001 MeV emitted by ^{234 mPa} in gamma ray measurement of a gypsum board that does not contain ²³⁸ U according to an embodiment of the present invention. The 1.001 MeV peak is hardly recognized in the corrected data spectrum obtained by subtracting the normal probability distribution that simulates the calculated actual peaks of 0.934 MeV and 1.120 MeV from the raw data spectrum. , ^238U can be confirmed to not exist.

FIG. 13 is a diagram showing the discrimination results of a peak of 1.001 MeV emitted by ^{234 mPa} in gamma ray measurement of a natural uranium ore containing ²³⁸ U according to an embodiment of the present invention. In the spectrum of the corrected data obtained by subtracting the calculated normal probability distribution that simulates the actual peaks of 0.934 MeV and 1.120 MeV from the spectrum of the raw data, a peak of 1.001 MeV is recognized, ²³⁸ It can be confirmed that U exists. It is possible to calculate the radioactive concentration of ²³⁸ U from the 1.001 MeV peak (count rate) and evaluate whether the value is below the detection limit (about 0.05 to 0.1 Bq/g). can. Thereby, the concentration of "purified uranium" can be calculated by analyzing the spectrum obtained by subtracting the "NORM" of uranium and thorium for each measurement target.

Embodiments of the present invention have been described with reference to the drawings. However, the present invention is not limited to these embodiments. The present invention can be implemented with various improvements, modifications, and variations based on the knowledge of those skilled in the art without departing from the spirit thereof. For example, in the embodiments of the present invention, uranium is assumed as the nuclear fuel material, etc., but the measurement method and calculation method are also applicable to other nuclear fuel materials, etc., and they can be applied to the invention. It is not excluded from the scope.

1 Measurement container 2 Radiation shield 3 Radiation detector 4 Detection section 5 Heavy equipment 6 Handle

Claims (5)

A measurement method for determining the radioactivity of purified uranium in the coexistence of purified uranium and uranium derived from the natural world in waste generated from facilities that handle nuclear fuel materials or nuclear source materials, comprising:
(A) preparing a measurement container having an opening, a bottom, and sides of a predetermined depth;
(B) arranging a NaI (Tl) scintillation radiation detector outside the side of the measurement container;
(C) measuring a γ-ray spectrum from the waste contained in the measurement container with the radiation detector;
(D) From the peak at 1.001 MeV after subtracting the background from the γ-ray spectrum, the peaks at 0.934 MeV and 1.120 MeV of ²¹⁴ Bi, which are adjacent and interfering, are subtracted, and the peaks generated in the decay process of ²³⁸ U are calculated. a step of discriminating a 1.001 MeV peak emitted by ^{234 m} Pa;
(E) A measuring method comprising the step of calculating the radioactivity of the purified uranium from the discriminated peak of 1.001 MeV. The step (D) of discriminating the 1.001 MeV peak,
(D1) subtracting a thorium series standard spectrum corrected according to the γ-ray measurement in the measurement container from the γ-ray spectrum to obtain a corrected γ-ray spectrum;
(D2) Subtracting the 0.934 MeV and 1.120 MeV peaks from the 1.001 MeV peak in the corrected γ-ray spectrum gives the 1.001 MeV peak emitted by ²³⁴ mPa generated in the decay process of ²³⁸ U. The measuring method according to claim 1, comprising the step of discriminating. The step (D1) of obtaining the corrected γ-ray spectrum includes:
(D10) measuring γ-rays from a predetermined thorium-series γ-ray source and subtracting the background to obtain a thorium-series spectrum;
(D20) performing density correction on the obtained thorium series spectrum in the measurement container to obtain a thorium series standard spectrum;
The measuring method according to claim 2, comprising: The step (D) of discriminating the 1.001 MeV peak,
(D3) calculating the standard deviation (σ) when fitting the total absorption peak of each energy of the spectrum data obtained by gamma ray measurement of the gamma ray standard source to a normal probability distribution;
(D4) Calculate the normal probability distribution (standard deviation (σ) at each peak) that simulates the actual interfering peaks of 0.934 MeV and 1.120 MeV from the net counts determined from 1.765 MeV of ²¹⁴ Bi. The process of
(D5) From the peak at 1.001 MeV in the corrected γ-ray spectrum, the calculated normal probability distribution simulating the actual peaks at 0.934 MeV and 1.120 MeV is subtracted to obtain the peak at 1.001 MeV. The measuring method according to claim 2, comprising the steps of: Step (C) of measuring the γ-ray spectrum from the waste,
(C1) Different KCl aqueous solutions were placed in the measurement container, and the standard efficiency (η ^E1.461 , unit: cps/ ₍ photon/cm ³ )))
(C2) determining the energy relative efficiency (Rη) of the NaI (Tl) scintillation radiation detector for γ rays;
(C3) using a cylindrical volumetric radiation source model to obtain a density correction coefficient (D _W · D _C ) for the standard efficiency when the density changes for water and concrete;
(C4) Measurement efficiency ηE _x (cps/(photon/cm ³ )) in the measurement container, standard efficiency (η _E1.461 ), energy relative efficiency (Rη), and density correction coefficient D _W · D _C Using the following formula η _Ex = η _E1.461 ×Rη×D _W・D _C
The measuring method according to any one of claims 1 to 4, comprising the step of determining from.

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