JP7266342B1 - Measurement method of radioactivity of uranium - Google Patents

Abstract

A method for measuring the radioactivity of refined uranium in the presence of refined uranium and NORM uranium in waste from a facility that handles nuclear fuel material or nuclear source material. The measuring method of the present invention includes the steps of: (A) providing a measurement container having an opening, a bottom, and sides of a predetermined depth; arranging a NaI(Tl) scintillation radiation detector; (C) measuring a γ-ray spectrum from the waste contained in the measurement container with the NaI(Tl) scintillation radiation detector; From the peak at 1.001 MeV after background subtraction from the line spectrum, subtracting the neighboring and interfering 214Bi peaks at 0.934 MeV and 1.120 MeV, the 234 mPa generated during the decay process of 238 U releases 1. and (E) calculating the radioactivity of purified uranium from the discriminated 1.001 MeV peak. [Selection drawing] Fig. 1

Description

TECHNICAL FIELD The present invention generally relates to a method for measuring radioactivity of uranium, and more particularly to a method for measuring radioactivity of uranium in waste generated from nuclear facilities and the like, and more specifically to the waste The present invention relates to a method for measuring the radioactivity of purified uranium in order to discriminate between purified uranium that may be contained therein and natural 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, non-radioactive waste (hereafter referred to as "NR waste") ) also occurs in large quantities. Regarding this NR waste, the Ministry of Economy, Trade and Industry has issued guidelines (instructions) for its handling (Non-Patent Document 1), and it is necessary to judge and manage it as waste according to its contents. For example, materials installed in controlled areas and items used in controlled areas that may be judged to be NR waste are required to undergo radiation measurement evaluation just in case.

In addition, concrete waste, etc., generated during the dismantling of buildings such as nuclear facilities, was not contaminated in the measurement of the primary decontamination, and if no problems were pointed out in the confirmation inspection by the Ministry of Education, Culture, Sports, Science and Technology after the primary decontamination, the building The controlled area was lifted, and it was outside the regulatory framework of the Nuclear Reactor Regulation Law (Reactor Regulation Law), and was able to be disposed of as waste from building dismantling (industrial waste). However, if the building demolition waste such as concrete waste is managed within the regulatory framework of the Reactor Regulation Law, it is required to conduct radiation measurement and evaluation just in case for security and safety. .

In addition, regarding items (including waste) taken out of the controlled area, other regulations, such as the Ordinance on Prevention of Ionizing Radiation Hazards (Ministry of Labor Ordinance No. 41, September 30, 1972), must be checked to ensure that there is no contamination. It is necessary to carry out after confirming.

Judgment as to whether or not a facility that handles uranium, which is one of nuclear facilities, falls under NR waste (hereinafter also referred to as “NR judgment”) shall It is supposed to be judged based on the fact that there is no clear history. The measurement performed when determining NR is to measure the total uranium in the waste to make sure there is no contamination. Confirmation of the absence of contamination can be based, for example, on the fact that uranium is below the theoretical detection limit.

However, this total uranium includes not only refined uranium used in research, but also uranium derived from the natural world (NORM: Naturally Occurring Radioactive Materials, hereinafter referred to as "NORM uranium") contained in raw materials for civil engineering materials such as concrete, soil, and abrasives. ) may be mixed. At that time, if NORM uranium is detected even if there is no residue of refined uranium, it will be necessary to apply either uranium recovery in the United States or domestic burial disposal, which will increase the processing amount and disposal cost. There is a risk of

In addition, just to be sure about refined uranium contamination, it is not possible to determine the presence or absence of residual refined uranium based only on the measurement result of total uranium. However, if it is possible to distinguish refined uranium from NORM uranium and confirm that refined uranium is below the detection limit (approximately 0.05 to 0.1 Bq/g), it is possible to show that there is no residual contamination from refined uranium. This is effective in terms of safety management at the facility. In addition, just to make sure that there is no generated furan other than NORM, measurements were taken, and by putting it out of the framework of the Reactor Regulation Law, it became possible to dispose of it as general waste, the same as the building dismantled after the controlled area was lifted. Become.

The above-described building dismantling wastes such as concrete scraps are usually stored in storage containers such as drums, and chemical analysis can be performed after sampling from the storage containers. However, in that case, ^ICP -MASS analysis of uranium isotopes ( ²³⁸ U and ²³⁵ U) is usually performed ^as a direct measurement. cannot be identified. In addition, it is possible to use a method of collecting samples from the storage container and analyzing them with a NaI (Tl) scintillation radiation detector (hereinafter also referred to simply as a NaI radiation detector) or a Ge semiconductor radiation detector. . However, in that 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.

NORM uranium includes those in permanent equilibrium with uranium as the parent nuclide, those with refined uranium as the starting nuclide, and those in permanent equilibrium with ²²⁶ Ra as the starting nuclide by some manipulation. is assumed. Approximately the same spectrum is obtained when each gamma-ray spectrum measurement is performed with the NaI radiation detector. The difference is the presence or absence of a slight gamma ray (1.001 MeV) emitted from ²³⁴ Th and ^{234 m} Pa 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 spectrum of ^{the 234 m} Pa gamma ray (1.001 MeV).

Patent Document 1 describes a method for nondestructively ^{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 intensity of 1.001 MeV γ-rays emitted by Pa from the γ-ray spectrum measured by a NaI radiation detector is disclosed.

However, the method of Patent Document 1 detects the amount of uranium in the above-mentioned building demolition waste, which is not a target for radioactive waste such as concrete waste, especially a trace amount of purified uranium in the coexistence of purified uranium and NORM uranium. not the way. In addition, in the method of Patent Document 1, first, calibration radioactive waste with a known uranium content is prepared based on the uranium analysis value obtained by the destruction measurement of radioactive waste, and the calibration radioactive waste of the present invention is used as the calibration radioactive waste. A calibration curve is created by applying the non-destructive measurement method of (1), and the amount of uranium is calculated using the calibration curve.

Japanese Patent Laid-Open No. 5-340861

Handling of "Non-Radioactive Waste" at Nuclear Facilities (Instruction), Nuclear and Industrial Safety Agency, Ministry of Economy, Trade and Industry, NISA-111a-08-1, May 27, 2008

SUMMARY OF THE INVENTION An object of the present invention is to provide a method for measuring the radioactivity of refined uranium in the presence of both refined uranium and NORM uranium in waste 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 presence of purified uranium and naturally occurring uranium in waste from a facility that handles nuclear fuel material or nuclear source material. The measurement method is
(A) providing a measurement vessel with an opening, a bottom and sides of a predetermined depth;
(B) positioning a NaI(Tl) scintillation radiation detector outside the sides of the measurement container;
(C) a step of measuring the γ-ray spectrum from the waste contained in the measurement container with a NaI (Tl) scintillation radiation detector;
(D) The peak at 1.001 MeV after background subtraction from the gamma-ray spectrum, minus the adjacent and interfering ²¹⁴ Bi peaks at 0.934 MeV and 1.120 MeV, generated during the decay process of ²³⁸ U. discriminating the 1.001 MeV peak emitted by ^{234 mPa} ;
(E) calculating the radioactivity of the 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 comprises:
(D1) obtaining a corrected γ-ray spectrum by subtracting a thorium-series standard spectrum corrected in accordance with 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 peak at 1.001 MeV emitted by ^{234 m} Pa generated during the decay process of ²³⁸ U. and

In one aspect of the present invention, the step (D1) of obtaining the corrected γ-ray spectrum comprises
(D10) measuring gamma rays from a predetermined gamma-ray source of the thorium series 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 comprises:
(D3) A step of obtaining the standard deviation (σ) when all absorption peaks of each energy of spectral data obtained by gamma-ray measurement of a gamma-ray standard radiation source are fitted to a normal probability distribution;
(D4) From the net counts determined from 1.765 MeV (15.4%) of ²¹⁴ Bi, the interfering real peaks at 0.934 MeV (3.03%), 1.120 MeV (15.1%) A step of calculating a simulated normal probability distribution (standard deviation (σ) at each peak);
(D5) A normal probability distribution simulating the actual peaks at 0.934 MeV (3.03%) and 1.120 MeV (15.1%) calculated from the peak at 1.001 MeV in the corrected γ-ray spectrum. subtracting to obtain a peak at 1.001 MeV (0.84%).

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

It is a figure which shows the basic flow of the measuring method of one Embodiment of this invention. It is a figure which shows the arrangement|positioning of the measurement container and radiation measuring device of one Embodiment of this invention. It is a figure which shows the calculation flow of the measurement efficiency of one Embodiment of this invention. FIG. 4 is a diagram showing density correction factors for one embodiment of the present invention; FIG. 4 shows the measured efficiency of one embodiment of the present invention; FIG. 4 is a diagram showing a processing flow of a γ-ray spectrum according to one embodiment of the present invention; FIG. 3 shows a thorium series standard spectrum of one embodiment of the present invention; FIG. 3 shows a uranium series standard spectrum of one embodiment of the present invention; FIG. 4 shows a process flow for discriminating the 1.001 MeV peak emitted by ^{234 m} 2 of one embodiment of the present invention. FIG. 4 is a diagram showing the standard deviation (σ) when all absorption peaks of each energy of γ-ray spectrum data are fitted to a normal probability distribution according to one embodiment of the present invention; FIG. 4 is a diagram showing normal probability distributions of 0.934 MeV (3.03%) and 1.120 MeV (15.1%) calculated according to an embodiment of the present invention; FIG. 10 shows the discrimination result of the 1.001 MeV peak emitted by 234 ^{m 2} Pa in the γ-ray measurement of the gypsum board not containing ²³⁸ U according to one embodiment of the present invention. FIG. 10 is a diagram showing the discrimination result of the 1.001 MeV peak emitted by 234 ^{m 2} Pa in gamma-ray measurement of natural uranium ore containing ²³⁸ U according to one embodiment of the present invention;

An embodiment of the present invention will be described with reference to the drawings. FIG. 1 is a diagram showing a basic flow of an inspection method according to one embodiment of the present invention. In step S1, a measurement vessel is provided that has an opening, a bottom and sides of a predetermined depth. In step S2, a NaI(Tl) scintillation radiation detector is placed outside the side of the measurement container. Step S2 may be performed simultaneously with 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, from the peak at 1.001 MeV after background subtraction from the gamma-ray spectrum, the neighboring and interfering peaks of ²¹⁴ Bi are subtracted at 0.934 MeV and 1.120 MeV, resulting in the decay process of ²³⁸ U. discriminate the 1.001 MeV peak emitted by ^{234 mPa} . In step S4, the radioactivity of the 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 meter according to one embodiment of the present invention. FIG. 2 shows a cylindrical measuring container 1 and a radiation detector 3 inserted in a radiation shield 2 positioned outside the side wall of the measuring container 1 . The measurement container 1 includes a cylindrical container with a substantially circular or elliptical opening. The measurement container 1 is, for example, a drum made of metal or resin. A measurement container 1 contains an aqueous solution of potassium chloride (KCl) to be measured, powdery waste (NR waste), shredded waste, compressible waste, flexible waste, and the like. Filling (storage) is possible.

The radiation shield 2 incorporates a heavy metal such as lead, and shields the radiation detector 3 from entering the radiation detector 3 from the outside as a background. One side wall of the radiation shield 2 is connected to the movable portion of the heavy machinery 5 . The radiation shield 2 has a cylindrical shape with a circular or semicircular opening, through which the radiation detector 3 can be taken in and out. By operating the handle 6 of the heavy machine 5 , the radiation shield 2 can be moved up and down along the side wall of the measurement container 1 . Thereby, the measurement position of the radiation detector 3 inside the radiation shield 2 can 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 can be arranged, or two or more radiation detectors 3 can be arranged within an elongated radiation shield 2 . The radiation detector 3 incorporates a detection section 4 at its tip. As the radiation detector 3, for example, a Na(Tl) scintillation detector having a crystal size of Φ63×160 mm in the detection part can be used. The crystal size is about three times as large as 63×63, and it can be expected that the detection accuracy will increase to the accuracy corresponding to NR measurement.

The radiation detector 3 is connected to a spectrum analyzer (not shown) and a computing device (not shown) via a communication cable (not shown). A communication cable is, for example, a GP-IB cable, a USB cable, a LAN cable, or the like. The spectrum analyzer measures the spectrum of γ-rays from the measurement object inside the measurement container 1 detected by the radiation detector 3 . The arithmetic unit calculates various parameters (measurement efficiency, radioactivity (density), etc.) for obtaining the radioactivity of uranium in the measurement object from the measured γ-ray spectrum. Computing devices include, for example, personal computers (PCs).

The waste that is filled in the measurement container 1 means the waste that is the target of NR waste generated in facilities that handle nuclear fuel materials and the like. The waste includes powdery waste, shredded waste, compressible waste, flexible waste, etc. that can be filled in the measurement container 1 . The powdery state in the powdery waste can include not only fine powder such as sand grains and rice grains, but also sizes of several millimeters (<about 10 mm). Waste can include, for example, materials such as flakes of paint and rusted iron, metal, pebbles, concrete pieces, wood, paper, plastic, poly sheet, vinyl, rubber, and the like.

Nuclear fuel materials and the like include uranium, thorium, and their compounds, which are α-, β-, and γ-ray sources. In addition to nuclear facilities, facilities that handle nuclear fuel materials include facilities that are not directly related to nuclear power, such as facilities that handle refined uranium, manufacturing facilities such as chemical plants that use uranium as a catalyst, and processing facilities. be Nuclear facilities include, for example, refining facilities, nuclear reactor facilities, and reprocessing facilities described in Non-Patent Document 1. Furthermore, NR waste refers to what is basically judged according to the guidelines in Non-Patent Document 1, but in addition to nuclear facilities, facilities that generate target waste are directly related to nuclear power as described above. Chemical factories, etc., which are not related to each other, are also included.

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

FIG. 3 is a diagram showing a flow of calculating measurement efficiency in consideration of self-absorption due to differences in the weight and material of the object to be measured when a drum can is used as the measurement container 1 . In step S10, different KCl aqueous solutions are placed in the measurement container 1 (drum) ^, and the total absorption peak area count rate (η _E1.461 , unit: cps/(photon /cm ³ ))), and this is defined as the standard efficiency (density: 1, material: water, E _γ : 1.461 MeV).

Table 1 shows information on the concentrations of the KCl aqueous solutions used in the tests. 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 calculated reference efficiency, that is, the total absorption peak area count rate (η _E1.461 , unit: cps/(photon/cm ³ )) of γ-ray of 1.461 MeV.

In step S11, the relative energy efficiency (R _η ) with respect to 1.461 MeV γ-rays at the NaI(Tl) scintillation radiation detector used is determined. Table 3 shows the relative energy efficiencies (R _η ) obtained.

In step S12, the cylindrical volume radiation source model is used to calculate the density correction coefficients (D _W ·D _C ) with respect to the standard efficiency when the densities of water and concrete change. Calculations are based on, for example, the cylindrical radiation source calculation method in the 1968 Engineering Compendium on Radiation Shielding IAEA. FIG. 4 shows the calculation results of the material and the density correction coefficient with a density of 1, a γ-ray energy of 1.461 MeV, and a case where water is used as the material of the contained object as a standard. The calculation results are represented 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 measured efficiency ηE _x (cps/(photons/cm ³ )) in the measurement container (drum) is calculated using the standard efficiency (η _E1.461 ), the relative energy efficiency (R _η ), and the density correction factor D _W・Using D _C , the following formula

η _Ex = η _E1.461 × Rη × _DW・D _C

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

A method for discriminating total absorption peaks at evaluated energies in overlapping gamma-ray spectra in a NaI(Tl) scintillation radiation detector is now described. When uranium- and thorium-series radioactive materials coexist, the gamma-ray spectrum obtained by measurement has overlapping broad peaks. When analyzing, it is necessary to remove these influences, separate the γ-rays of the energy to be evaluated into individual peaks, and calculate the areas. Therefore, it is necessary to evaluate according to the γ-ray spectrum processing flow of one embodiment of the present invention shown in FIG. In addition, when some energy peaks interfere even in a single series radioactive substance, data processing is performed according to the γ-ray spectrum processing flow of one embodiment of the present invention shown in FIG. 9 described later, and a single It is necessary to treat it as an energy peak.

In step S21 of the γ-ray spectrum processing flow of one embodiment of the present invention in FIG. 6, γ-ray spectra in radiation equilibrium are measured using uranium and thorium series radiation sources that can ignore self-absorption, Background subtraction provides standard spectra for uranium and thorium. FIG. 7 shows the thorium series standard spectrum obtained. FIG. 8 shows the obtained uranium series standard spectrum. In step S21, the standard spectrum of thorium is matched to the 2.615 MeV peak of ²⁰⁸ Tl in the γ-ray spectrum of the waste, and a thorium-series standard spectrum is calculated with a density correction factor added at the same time.

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

Next, a method for evaluating the concentration of refined uranium when refined uranium is mixed in the uranium series will be described. Among the radioactive materials treated as NORM, the uranium series includes those in permanent equilibrium with uranium as the parent, those with purified uranium as the starting nuclide, and those with ²²⁶ Ra as the starting parent nuclide. "Things in a state" etc. are assumed. To estimate the radioactivity of the purified ²³⁸ U, 0.934 MeV (3.03%) of adjacent and interfering ²¹⁴ Bi and 1.934 MeV (3.03%) of 214 Bi were extracted from the corrected gamma-ray spectra obtained by removing the thorium series peaks. The 120 MeV (15.1%) peak must be subtracted to discriminate the 1.001 MeV (0.84%) peak emitted by ^{234 mPa} generated during the decay process of ²³⁸ U. It should be noted that similar processing can be performed for peaks at other energies. FIG. 9 shows such a processing flow.

In step S30 of FIG. 9, the standard deviation (σ) is obtained when all absorption peaks of each energy of the line spectrum obtained by the standard radiation source are fitted to the normal probability distribution. All absorption peaks of the γ-ray spectrum obtained using a standard point source of ¹³⁷ Cs (661 keV) and ⁶⁰ Co (1173 keV, 1333 keV) were clipped by trapezoidal approximation of the background and fitted with a normal probability distribution by the least squares method. FIG. 10 shows the standard deviation (.sigma., representing the broadening of the peak) obtained when

In step S31, relative efficiencies of target γ-rays (1.765 MeV, 0.934 MeV, 1.120 MeV) are obtained from the weight and material of the waste in the measurement container. In step S32, the interfering actual peaks at 0.934 MeV (3.03%), 1.120 MeV (15.1%), 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) simulating

Specifically, for example, a normal probability distribution is calculated as follows.
(1) The peak area (number of counts) C of 1.765 MeV of ²¹⁴ Bi is corrected by the γ-ray emission rate ratio of ²¹⁴ Bi. That is, in the peak area (number of counts) 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 ratio) = 15.1%/15.4% to get _C1 and _C2 .
(2) Perform correction based on the γ-ray detection efficiency ratio of the NaI radiation detector. The gamma-ray detection efficiency ratio is calculated in advance from the calibration radiation source measurement results. Specifically, the γ-ray detection efficiency ratio (1.48 _{(0.934 MeV)} /0.86 _{(1.765 MeV)} , 1.39 _{(1.120 MeV)} /0.86 _{(1 .765 MeV)} ) is multiplied by C ₁ and C ₂ obtained in (1) to obtain C ₁₁ and C ₂₂ .
(3) Perform correction based on the γ-ray self-absorption rate ratio according to the density of the waste in the measurement container, which is the object of measurement. The γ-ray self-absorption rate ratio is calculated in advance by cylindrical radiation 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 .120 MeV)} /0.897 _{(1.765 MeV)} ) 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 simulated peak areas (number of counts) of 0.935 MeV and 1.120 MeV.

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

In step S33, 0.934 MeV (3.03%), 1.120 MeV (15.1 %) is subtracted to obtain a peak at 1.001 MeV (0.84%).

From the measured spectra of natural uranium ore containing ²³⁸ U and gypsum board, in which nuclides above ²²⁶ Ra are in radial equilibrium with prior Ge semiconductor detectors and which are known to contain no ²³⁸ U, An experiment was conducted to obtain a peak of 1.001 MeV by subtracting a normal probability distribution simulating the actual peaks of 0.934 MeV and 1.120 MeV calculated in step S32.

FIG. 12 is a diagram showing the discrimination result of the peak of 1.001 MeV emitted by ^{234 m} 2 in γ-ray measurement of gypsum board not containing ²³⁸ U according to one embodiment of the present invention. The 1.001 MeV peak is hardly recognized in the corrected data spectrum obtained by subtracting the calculated normal probability distribution that simulates the actual peaks at 0.934 MeV and 1.120 MeV from the raw data spectrum. , ²³⁸ U does not exist.

FIG. 13 is a diagram showing the discrimination result of the 1.001 MeV peak emitted by ^{234 m} 2 in γ-ray measurement of natural uranium ore containing ²³⁸ U according to one embodiment of the present invention. In the spectrum of the corrected data obtained by subtracting the calculated normal probability distributions simulating the actual peaks of 0.934 MeV and 1.120 MeV from the raw data spectrum, a peak of ^1.001 MeV is recognized. It can be confirmed that U exists. It is possible to calculate the radioactivity concentration of ²³⁸ U from the 1.001 MeV peak (count rate) and evaluate whether the value is below the detection limit value (about 0.05 to 0.1 Bq / g). can. As a result, the "refined uranium" concentration can be calculated by analyzing the spectrum from which uranium and thorium, which are "NORM" for each measurement object, are subtracted.

Embodiments of the present invention have been described with reference to the drawings. However, the 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 of the invention. For example, in the embodiments of the present invention, uranium is assumed as a nuclear fuel material, etc., but the measurement method and calculation method are applicable to other nuclear fuel materials, etc., and they are applicable to the invention. It is not excluded from the target.

1 Measurement Container 2 Radiation Shield 3 Radiation Detector 4 Detector 5 Heavy Equipment 6 Handle

Claims (4)

A measurement method for determining the radioactivity of purified uranium in the presence of purified uranium and naturally occurring uranium in waste from facilities that handle nuclear fuel material or nuclear source material, comprising:
(A) providing a measurement vessel with an opening, a bottom and sides of a predetermined depth;
(B) placing a NaI(Tl) scintillation radiation detector outside the sides of the measurement container;
(C) measuring a γ-ray spectrum from the waste contained in the measurement container with the radiation detector;
(D) The peak at 1.001 MeV after background subtraction from the gamma-ray spectrum, minus the adjacent and interfering ²¹⁴ Bi peaks at 0.934 MeV and 1.120 MeV, generated during the decay process of ²³⁸ U. discriminating the 1.001 MeV peak emitted by ^{the 234 mPa} that
(E) calculating the radioactivity of the purified uranium from the discriminated 1.001 MeV peak ;
The step (D) of discriminating the 1.001 MeV peak is
(D1) obtaining a corrected γ-ray spectrum by subtracting a thorium series standard spectrum corrected in accordance with the γ-ray measurement in the measurement vessel from the γ-ray spectrum;
(D2) A 1.001 MeV peak emitted by ^{234 m} Pa generated in the decay process of ²³⁸ U by subtracting the 0.934 MeV and 1.120 MeV peaks from the 1.001 MeV peak in the corrected gamma-ray spectrum. A method of measurement comprising the step of discriminating between The step (D1) of obtaining the corrected γ-ray spectrum is
(D10) measuring gamma rays from a predetermined gamma-ray source of the thorium series and subtracting the background to obtain a thorium series spectrum;
(D20) The measurement method according to claim 1 , comprising the step of performing density correction in the measurement container on the obtained thorium series spectrum to obtain a thorium series standard spectrum. The step (D) of discriminating the 1.001 MeV peak is
(D3) A step of obtaining the standard deviation (σ) when all absorption peaks of each energy of spectral data obtained by gamma-ray measurement of a gamma-ray standard radiation source are fitted to a normal probability distribution;
(D4) Calculate the normal probability distribution (standard deviation (σ) at each peak) simulating the actual interfering peaks of 0.934 MeV and 1.120 MeV from the net counts obtained from 1.765 MeV of ²¹⁴ Bi. and
(D5) Subtract the calculated normal probability distribution simulating the actual peaks at 0.934 MeV and 1.120 MeV from the peak at 1.001 MeV in the corrected γ-ray spectrum to obtain the peak at 1.001 MeV. The method of claim 1 , comprising the steps of: The step (C) of measuring a γ-ray spectrum from the waste is
(C1) ^Different KCl aqueous solutions are placed in the measurement container, and the standard efficiency (η _E1.461 , unit: cps/(photon/cm ³ )), and
(C2) determining the relative energy efficiency (Rη) of the NaI(Tl) scintillation radiation detector for gamma rays;
(C3) Using a cylindrical volume radiation source model, for water and concrete, a step of obtaining a density correction coefficient (D _W · D _C ) with respect to the standard efficiency when the density changes;
(C4) The measurement efficiency ηE _x (cps/(photons/cm ³ )) in the measurement container, the reference efficiency (η _E1.461 ), the energy relative efficiency (Rη), and the density correction factor D _W · D _C using the following formula η _Ex = η _E1.461 × Rη × D _W · D _C
The method of claim 1 , comprising the step of determining from

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